bluuebunny/crossref_metadata_2025_split huggingface dataset:
A clean nuclear blasting technique for the excavation of a new isthmus canal in Central America
Authors: Seifritz W. 
Published: December 2013

Abstract
Based an new insights into the physics of thermonuclear burn waves a novel nuclear blasting technique is proposed for the excavation of a new Isthmus canal in Central America. It allows the spatial decoupling of a small fission trigger from the main DD fusion charge in the geometrical form of a cylinder with an extremely high length-to-diameter ratio in such a way that absolutely no fission products can escape into the atmosphere. A clean nuclear blasting technique could thus be made available.
Practical applications of the exemption principle in radiation protection
Authors: Wrixon A. D. 
Published: December 2013

Abstract
The concept of exemption has been widely used in radiological protection. It is part of the regulator's arsenal for applying legislative requirements in a graded fashion, to avoid the expenditure of effort on situations where the return in terms of improvement in protection would otherwise be trivial. Nevertheless, it still remains a controversial matter. It is not so much the dosimetric criteria for exemption that cause debate; it is more the way in which the concept is used; the scenarios employed to calculate derived activity concentrations, and its relationship to the somewhat analogous concept of exclusion. A particular problem has been the full integration into the legislative system of protection of exposures to radiation from sources of natural origin and the degree to which exemption is a relevant concept for dealing with such situations. The purpose of this paper is to attempt to provide some clarity on the two concepts and their practical implementation with a view to encouraging international harmonization and avoiding further unnecessary debate.
Determination of the replacement cooling tower capability at the ETRR-2 research reactor
Authors: El-Morshdy S. El-Din 
Published: December 2013

Abstract
The ETRR-2 replacement cooling tower capability has been evaluated by the thermal acceptance test performed in June 2003. All instruments used were calibrated prior to the test. The measured data are collected at regular intervals in accordance with the acceptance test code for water cooling towers of the Cooling Tower Institute recommendations. Both the characteristic curve and the performance curve methods were used to evaluate the tower capability. The test results yield a tower capability of about 105% and so the tower is thermally accepted.
Brennstabauslegung und Brennstabmodellierung – Teil 1
Authors: von Collani Elart , Baur Karl 
Published: December 2013

Kurzfassung
Die Brennstabauslegung ist ein wesentlicher Aspekt für den sicheren und wirtschaftlichen Betrieb von Kernkraftwerken. Die langjährige Betriebserfahrung zeigt, dass der Sicherheitsstandard der Brennstäbe sehr hoch ist und die international anerkannten Schutzziele eingehalten werden. Zwischenzeitlich haben wirtschaftliche Überlegungen zu einer Erhöhung der mittleren Brennelementabbrände von 35 MWd/kgSM auf heute etwa 55 MWd/kgSM geführt. Neben der besseren Wirtschaftlichkeit wird dadurch im Sinne des Umweltschutzes der Kernbrennstoffbedarf reduziert und die Anzahl der zu entsorgenden Brennelemente verringert. Allerdings bedeutet jede weitere Abbranderhöhung auch zusätzliche Anforderungen an die Brennstäbe, insbesondere an die Hüllrohrmaterialien, wodurch sich erneut die Notwendigkeit des Nachweises der Brennstabintegrität ergibt. Im Vordergrund dieser Nachweise steht seit geraumer Zeit der Übergang von deterministischen zu statistischen Auslegungsmethoden, mit dem Ziel durch Berücksichtigung der vorhandenen Variabilität der Vorgänge im Reaktorkern zu einer wirklichkeitsnäheren Beschreibung zu gelangen. Dadurch können zuverlässigere und genauere Vorhersagen über das Verhalten der Brennstäbe gemacht werden, die die Grundlage für eine effektive Brennstabauslegung und den Nachweis der Brennstabintegrität bilden. Dieser Beitrag beschäftigt sich mit der Frage, inwiefern die bestehenden Modelle aus stochastischer Sicht die Variabilität berücksichtigen und zeigt, dass ein stochastisches Modell neue Chancen für zuverlässige Nachweise bietet. Das vorgeschlagene stochastische Modell behebt gewisse Schwächen der bisherigen Vorgehensweisen, setzt alle vorhandenen Erkenntnisse direkt um und macht die stets vorhandene Ignoranz sichtbar. Zusätzlich werden im Beitrag einige, insbesondere in der Bundesrepublik Deutschland aufgeworfene Fragen, kritisch betrachtet. Der Beitrag besteht aus zwei Teilen. Im ersten Teil werden die benötigten Grundlagen gelegt, während im zweiten Teil auf die mehr technischen Fragen eingegangen wird.
Decommissioning strategy selection
Authors: Warnecke E. 
Published: December 2013

Abstract
At the end of their useful life nuclear facilities have to be decommissioned. The strategy selection on how to decommission a facility is a highly important decision at the very beginning of decommissioning planning. Basically, a facility may be subject to (a) immediate dismantling; (b) deferred dismantling after a period of “safe enclosure” or (c) entombment where a facility is turned into a near surface disposal facility. The first two strategies are normally applied. The third one may be accepted in countries without significant nuclear activities and hence without disposal facilities for radioactive waste. A large number of factors has to be taken into account when a decision on the decommissioning strategy is being made. Many of the factors cannot be quantified. They may be qualitative or subject to public opinion which may change with time. At present, a trend can be observed towards immediate dismantling of nuclear facilities, mainly because it is associated with less uncertainty, less local impact, a better public acceptance, and the availability of operational expertise and know how. A detailed evaluation of the various factors relevant to strategy selection and a few examples showing the situation regarding decommissioning strategy in a number of selected countries are presented in the following article.
Decommissioning technologies, including recent developments and special features of the dismantling of nuclear research and prototype facilities
Authors: Bach Fr.-W. , Pfeifer W. , Versemann R. , Wilk P. , Valencia L. , Eisenmann B. , Hammer G. 
Published: December 2013

Abstract
A variety of technologies is available for the decontamination and dismantling of nuclear facilities. All over the world, these technologies as well as conditioning processes of decommissioning waste are further developed. So far, they have been mainly applied to the dismantling of research and prototype facilities. Dismantling of reactors of higher power has started with the nuclear power plants of Greifswald and Würgassen among others. The present paper shall present the state of the art of decontamination and dismantling techniques as well as of post-treatment processes of radioactive decommissioning waste. From the technical point of view, the most complex step is the remote dismantling of activated and highly contaminated components. The prototype facilities of Forschungszentrum Karlsruhe, which are currently being dismantled, cover all types of plants. Their status of dismantling and the experience gained from using decommissioning technologies that are also suited for power facilities shall be illustrated. Further developments and adaptations of individual techniques, e. g. thermal and mechanical cutting methods, as well as of the complete systems technology, including carrier system, manipulator system, and tools, are reasonable and indispensable for managing the variety of dismantling tasks and, not least, for reducing their operation times and costs of use.
The decommissioning of light water reactors – experience
Authors: Schiffer K. , Bacmeister G. U. , Adler J. 
Published: December 2013

Abstract
More than ten nuclear power plants in Germany have been decommissioned or are under dismantling. Objectives, strategies, preparatory work, licensing, technical dismantling and safety considerations are explained from the utility point of view, using the Stade (KKS) decommissioning as an example. Challenges with respect to characterisation, logistics and economy are discussed.
Decommissioning and dismantling of the Siemens fuel cycle facilities
Authors: Kindleben G. , Baumann R. , Faber P. 
Published: December 2013

Abstract
Four Siemens-owned German fuel cycle facilities are being decommissioned at the Hanau and Karlstein sites. Decommissioning of one of these facilities has already been completed. The facilities comprise three fuel fabrication plants and a facility for post-irradiation investigations of fuel assemblies. In each case the actual decommissioning phase was preceded by clean-out. Extensive preparatory investigations and planning, including licensing procedures, had been necessary. In some cases new measuring techniques and systems had to be developed and waste processing systems erected. In addition to providing a general survey of the different conditions and procedures involved, the MOX facility is used in this paper as an example for presenting these activities in greater detail. This includes the concept of immediate conditioning of radwaste for maintenance-free long-term interim storage (storage for several decades until a repository is available) in such a way as to also meet the requirements for shipment and final disposal without any need for further actions. A number of lessons have been learned and recommendations derived from the many years of decommissioning work at the four facilities.
Neutronic conceptual design of the ETRR-2 cold neutron source
Authors: Khalil M. Y. , Shaat M. K. , Abdelfattah A. Y. 
Published: December 2013

Abstract
The conceptual neutronic design of the cold neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was carried out using the MCNP code. A parametric analysis was also performed to choose the type and geometry of the moderator and the required CNS dimensions to maximize the cold neutron production. The moderator cell is a spherical annulus containing liquid hydrogen. The cold neutron gain and brightness are calculated together with the nuclear heat load of the CNS. The effects of void fraction in the moderator cell and the ortho:para ratios on cold neutron gain were calculated.
Simulation of the neutronic-thermal hydraulic stability behaviour of boiling water reactors
Authors: Gorzel A. H. , Henschel M. W. , Mrosk E. M. , Scheerer W. E. 
Published: December 2013

AbstractNeutronic-thermal hydraulic oscillations are a well known phenomenon in the operation of boiling water reactors. Despite the progresses that have been made concerning their simulation, the computational determination of the stability threshold in the power-flow map still suffers from inaccuracies. In this work, the decay ratio of neutronic-thermal hydraulic oscillations and hence the stability threshold was calculated with the SIMULATE-3K code developed by Studsvik Scandpower. It is known from experience that there is a higher probability for a less stable reactor operation in the case of one or more pumps tripped while the remaining pumps run at minimum speed. For the analysis of this state, the characteristics of the pump system were modelled carefully. It is shown that the stability behaviour is determined by the axial and radial power profile due to control rod manoeuvring and fuel burn-up and by the number of pumps running. Furthermore the calculations reveal the impacts of the insertion of shallow rods and of the control-rod group insertion on the decay ratio. A comparison of the calculations with experimental results of stability measurements shows that by a thorough analysis and proper modelling of the plant characteristics the decay ratios can be calculated with the SIMULATE-3K code.
Validation of the CFD code FLUENT by post test calculation of a ROCOM experiment with density driven coolant mixing
Authors: Schaffrath A. , Fischer K.-C. , Hahm T. , Wussow S. 
Published: December 2013

Abstract
During the last years one focus of German PWR safety analyses was boron dilution events with the potential of reactivity transients. Coolant with a low boron concentration could be collected in localized areas of the reactor coolant system e. g. by separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. The TüV NORD SysTec appraises in the framework of follower core assessments safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses shall demonstrate that boron dilution events cannot lead to recriticality of the core. Hence the boron concentration at the core inlet has to be determined. TüV NORD Sys Tec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events a validation against the ROCOM experiment T6655_21 with density driven coolant mixing was performed. This validation proves that FLUENT is able to determine the boron concentration in the reactor core with sufficient accuracy and to solve the safety issue, whether the core gets critical or not.
Production of 4He and 3He from tantalum and tungsten irradiated with nucleons at energies up to 1 GeV
Authors: Broeders C. H. M. , Konobeyev A. Yu. 
Published: December 2013

Abstract
The comparison of different methods of calculation and computer codes used to obtain helium isotope production cross-sections for nuclear reactions induced by nucleons at energies up to 1 GeV was performed. The cross-sections were calculated using the GNASH code, the modified ALICE code, the DISCA code and the CASCADE/INPE code. The evaluation of the helium production cross-section for tantalum and tungsten was performed using the results of model calculations, available experimental data and systematics predictions. The cross-sections were obtained for nuclear reactions induced by neutrons and protons at energies up to 1 GeV
Approximation method for the adjoint neutron spectrum in heterogeneous media
Authors: Suster L. C. , Martinez A. Senra , da Silva F. Carvalho 
Published: December 2013

Abstract
A method is proposed for the numerical and analytical calculations of the adjoint neutron spectrum in the resonance energy region, for narrow resonances, in order to reduce the numerical computations involved. The adjoint neutron spectrum of infinite cylindrical fuel cell is calculated according to a very simple analytical method with reasonable level of accuracy. The results obtained with this analytical method were compared to a reference solution obtained with a numerical solution of the neutron balance adjoint equations. The adjoint neutron spectrum is useful to determine the neutron resonance absorption, so that multigroup adjoint cross sections used by the adjoint diffusion equation can be obtained.
A fool proof non-proliferation nuclear reactor concept
Authors: Sefidvash F. , Seifritz W. 
Published: December 2013

AbstractThe non-proliferation characteristics of the Fixed Bed Nuclear Reactor (FBNR) is based on both, the extrinsic concept of sealing and the intrinsic concept of isotope denaturing. Its small spherical fuel elements are confined in a fuel chamber that can be sealed by the authorities for inspection at any time. Only the fuel chamber is needed to be transported from the fuel factory to the site and back. There is no possibility of neutron irradiation to any external fertile material. Isotopic denaturing of the fuel cycle either in the U-233/Th or Pu-239/U cycle increases the proliferation resistance substantially. Therefore, both concepts of “sealing” and “isotope denaturing” contribute to the fool proof non-proliferation characteristics of the proposed reactor. This contribution is focussing only on the won-proliferation aspect of the reactor.
Best estimate analysis of group distribution header blockage events in RBMK-1500 reactors
Authors: Kaliatka A. , Vaisnoras M. , Vileiniskis V. 
Published: December 2013

Abstract
The analysis of Group Distribution Header (GDH) blockage events shows that the integrity of several fuel channels could be violated. A best estimate analysis of GDH blockage events in the RBMK-1500 reactor (Large Channel Type Water-cooled Graphite-moderated Reactor) of the Ignalina Nuclear Power Plant is performed. Uncertainties of the initial and boundary conditions as well as ATHLET code models were taken into account during the simulation. The analysis is performed employing the code ATHLET Mod. 1.2 Cycle D and the package SUSA 3.5. In this paper the ATHLET code model of Ignalina NPP, event scenarios and results of the analysis are presented. Analysis results show that GDH blockage events do not lead to severe consequences.
Neutronic calculations for the new fuel configuration of the ETRR-1 research reactor
Authors: Aziz M. 
Published: December 2013

Abstract
Neutronic calculations were performed for the new loading configuration of the ETRR-1 research reactor. The MCNP three dimensions Monte Carlo code and the two dimensions CITATION code are used to model the reactor. The power and thermal flux distributions in the reactor core are calculated. The power peak factor and the effect of control rod insertion on both flux and power profiles in the reactor core are determined and analyzed. The partial and total control rods worth are calculated. It was found that the difference between MCNP and CITATION in power distributions is 4 to 8 % and for thermal flux ranges between 3 to 14 %.
The HN method for slab albedo problem for linearly anisotropic scattering
Authors: Bulut S. , Güleçyüz M. Ç. 
Published: December 2013

Abstract
The HN method is used to calculate the reflection and transmission coefficients for a parallel plane media with linearly anisotropic scattering. The method depends on the use of the angular distributions of the method of elementary solutions. The ortogonality relations of the singular eigenfunctions together with the values of the angular distributions at the boundaries of the given medium lead to the solution of the problem. It is shown that the method leads to concise equations and to fast converging numerical results.
CCF treatment in PSA: insights and recommendations from reviewing procedures
Authors: Spitzer C. 
Published: December 2013

Abstract
Probabilistic Safety Assessments (PSAs) in Germany have primarily been conducted in the framework of Periodic Safety Reviews (PSRs). For some time, PSAs following the recommendations given by the reviewer with the final statement in the 1st PSR have been submitted and reviewed, like the probabilistic analyses of internal fires as well as Low Power and Shutdown States. From the very beginning, the review of the adequateness of the common cause failures (CCF) treatment represented a key element in the respective regulatory procedures – given the large impact on the results depending on the CCF assessment. In this paper, our experiences and insights from the reviewing procedures in the course of time are described – also addressing background and objectives as well as working programme and results with reference to reviewer's CCF projects. Finally, possible ways and procedures are presented related to a further development in CCF treatment in PSAs according to practical requests.
Further development of the coupling model
Authors: Kreuser A. , Peschke J. , Stiller J. C. 
Published: December 2013

Abstract
Uncertainties arising from different sources have to be considered for the quantification of common cause failures (CCFs). At GRS a CCF model (coupling model) has been developed for the estimation of CCF probabilities. An essential feature of the coupling model is the consideration of these uncertainties by using Bayesian estimation methods. Experiences from applying the coupling model to CCF event data over several years and analyzing the results in detail has led to improvements in the application of the model. In this paper the improved methodology of the coupling model is presented. Special emphasis is given to the description of the sources of uncertainties which are considered in the coupling model and the mathematical methodology, how these uncertainties are represented and propagated through the model. In closing topics of future improvements of the coupling models are discussed.
The Process-Oriented Simulation (POS) model for common cause failures: recent progress
Authors: Berg H.-P. , Görtz R. , Schimetschka E. , Kesten J. 
Published: December 2013

AbstractA common-cause failure (CCF) model based on stochastic simulation has been developed to complement the established approaches and to overcome some of their shortcomings. Reflecting the models proximity to the CCF process it was called Process Oriented Simulation (POS) Model. In recent years, some progress has been made to render the POS model fit for practical applications comprising the development of parameter estimates and a number of test applications in areas where results were already available – especially from CCF benchmarks – and comparison can provide insights in strong and weak points of the different approaches. In this paper, a detailed description of the POS model is provided together with the approach to parameter estimation and representative test applications. It is concluded, that the POS model has a number of strengths – especially the feature to provide reasonable extrapolation to CCF groups with high degrees of redundancy – and thus a considerable potential to complement the insights obtained from existing modeling.
Uncertainty and sensitivity analysis of the CPS-CC voiding in the RBMK reactor Ignalina-2
Authors: Tonkunas A. , Pabarcius R. , Clemente M. , Listopadskis N. 
Published: December 2013

Abstract
The best-estimate accident analysis approach is widely used for studying safety issues. The introduction of nominal values implies the evaluation of solution uncertainty, therefore an uncertainty analysis is required. New results of an uncertainty and sensitivity analysis of the Control and Protection System Cooling Circuit (CPS-CC) voiding accident in an RBMK-1500 reactor core are presented here. A typical reactor core state of Ignalina NPP Unit2 before the implementation of the independent diverse shut-down system was chosen for the study. The analyzed reactivity initiated accident is one of the most severe ones in RBMK-1500 reactors causing a positive reactivity insertion of about 2 in the case of a full CPS-CC voiding. The simulation of the CPS-CC dynamic voiding process was conducted without taking into account an additional failure of the reactor equipment. The analysis was performed by employing the GRS uncertainty and sensitivity method based on the SUSA software package in combination with the 3D core model QUABOX/CUBBOX-HYCA, adjusted for reactors of RBMK type. The results of the analysis give indications for which parameters the uncertainty of knowledge should be reduced in order to reduce the uncertainty of results most effectively.