bluuebunny/crossref_metadata_2025 huggingface dataset:
EFNDT Working Group 5 – der Brückenschlag zwischen technischer und öffentlicher Sicherheit
Authors: Osterloh Kurt , Wrobel Norma , Kolkoori Sanjeevareddy
Published: December 2013

Kurzfassung

Die Arbeitsgruppe 5 der Europäischen Föderation für Zerstörungsfreie Prüfung (EFNDT Working Group 5 “Public Security and Safety”, EFNDT WG 5) verfolgt weiterhin das Ziel, organisatorisch weit auseinanderliegende Zuständigkeiten und Arbeitsgebiete zusammenzuführen, die weitgehend mit gleichen, zumindest ähnlichen Prinzipien und Methoden arbeiten. Auf der Suche nach neuen Wegen stoßen beide Seiten auf Gemeinsamkeiten. Dies gilt auch für neue Ansätze und Bewertungsmethoden von Techniken, mit denen Risiken vermindert werden sollen. Hierzu hat in Berlin ein Workshop im Oktober 2012 stattgefunden, auf dem parallele Projekte zur Container- und Frachtsicherheit zusammengeführt werden konnten. Herausgestellt wurde die Komplementarität der laufenden Projekte, gemeinsame Aspekte sollen künftig auch zusammen bearbeitet werden. Darüber hinaus wurden neue Methoden und Vorgehensweisen vorgestellt, aber auch die Sichtweise der Europäischen Kommission mit ihren Vorstellungen zu einem europäischen Markt für Sicherheitstechnik dargestellt. Angesprochen wurden bestehende und vorgeschlagene europaweite Regulierungen und das europäische Referenznetz zum Schutz kritischer Infrastrukturen (ERNCIP), welches ein Netzwerk von Laboratorien für den Test von Sicherheitstechnik zum Ziel hat. Auf technischem Sektor standen Methoden im Vordergrund, mit denen die Gesamtmenge eines Gefahrstoffes wie z. B. Sprengstoff erkannt werden kann, und nicht nur Spuren als indirekter Hinweis. Die Arbeitsgruppe konnte somit erstmalig ein neutrales Forum für parallel laufende Projekte zur öffentlichen Sicherheit zur Verfügung stellen. Die auf den Workshops 2010 bis 2012 in Berlin gehaltenen Vorträge sind auf der Webseite des EFNDT WG 5 im geschützten Bereich zu finden. Nicht alles davon ist geeignet, es der allgemeinen Öffentlichkeit zugänglich zu machen. Deshalb sind alle Vorträge aus den Workshops in einem passwortgeschützten Verzeichnis abgelegt. Die Erteilung des Passwortes erfolgt über den Convenor der Arbeitsgruppe durch die EFNDT. Eine Erweiterung dieser Struktur zu einem Austauschforum ist vorgeschlagen. Zur Teilnahme daran sind nicht nur Mitglieder in den Gesellschaften für zerstörungsfreie Prüfung eingeladen, sondern auch diejenigen, die in den Anwendungsgebieten zur öffentlichen Sicherheit tätig sind.
Einfluss der Auswertestrategie auf die Grenzformänderungskurve∗
Authors: Kuppert Andreas , Merklein Marion , Friebe Harald , Möller Theodor , Hotz Walter , Volk Wolfram
Published: December 2013

Kurzfassung

Die Grenzformänderungskurve stellt bis heute das bewährte und etablierte Verfahren zur Bestimmung der Versagensgrenze des Werkstoffes dar und kann für nahezu alle modernen Simulationsprogramme der Blechumformung als Eingangsparameter verwendet werden. Die Qualität der FLC ist entscheidend von der Definition des Beginns der Einschnürung während der Umformung abhängig und geht auf Arbeiten von Keeler [1] und Goodwin [2] zurück. In diesem Beitrag wird dargestellt, dass besonders bei Leichtbauwerkstoffen das standardisierte Verfahren nach ISO 12004-2 [3] Schwächen aufweisen kann, wodurch Abweichungen zwischen Simulationsergebnissen und realen Umformversuchen entstehen können. Die unterschiedlichen Ansätze zur Ermittlung der FLC werden gegenübergestellt. Im besonderen Fokus stehen dabei die zeitlichen Auswertestrategien, welche aktuell auch im Rahmen der Arbeitsgruppe „Erweiterung FLC ISO 12004“ diskutiert werden.
Microstructure and Mechanical Properties of HVOF Sprayed WC-Co/NiCrBSi, Cr3C2 Coatings on Al Alloys
Authors: Özel Serkan
Published: December 2013

In this study, WC-Co/NiCrBSi and Cr3C2 powders were coated on surface of Al alloys (AA7075-T651, AA5083, AA2024-T4) by using HVOF (High Velocity Oxy Fuel) process. The coated samples were characterized by means of a scanning electron microscope (SEM), energy dispersive X-ray spectroscopy (EDX, XRD), hardness and bonding strength. These results showed that the maximum hardness and minimum surface roughness have been obtained at specimens coated with WC-Co. The bonding strengths of specimens used AA2024-T4 substrates were maximum. XRD study revealed that WC, W2C, C, Co, Cr3Ni2SiC, W5Si3, Cr7C3, Cr3C2, Cr, and Ni phases are present in the coatings.
Hochenergieradiografie – Erkennung von Einzelheiten in großen, komplexen Schichttiefen
Authors: Kolkoori Sanjeevareddy , Wrobel Norma , Osterloh Kurt , Redmer Bernhard , Deresch Andreas , Ewert Uwe
Published: December 2013

Kurzfassung

Zur Durchstrahlung großer, dichter Objekte reichen die Energien, die mit normalen Röntgenröhren erzielt werden können, ab gewissen Schichtdicken nicht mehr aus. Als hochenergetische Strahlenquellen stehen Kobalt-60 und Elektronenbeschleuniger zur Verfügung. Als Einschränkung der Hochenergieradiografie ist der geringere Kontrast besonders beim Vorhandensein leichterer Objekte umgeben von Körpern bestehend aus schweren Elementen anzusehen. Es ist daher zu erwarten, dass Objekte aus organischen Substanzen hinter dicken Schwermetallwänden oder in Bohrungen von Metallblöcken schwer zu erkennen sind. Im Unterschied zu einem Gammastrahler wie Kobalt-60 mit den beiden Spektrallinien um 1,3 MeV besitzt die Bremsstrahlung aus einem Beschleuniger einen wesentlichen Anteil an niederenergetischer Strahlung, der fließend zu höheren Energien übergeht. Es wird hier untersucht, welche Signaturen von leichten Materialien in einer Umgebung aus Schwermetall bis zu welcher Dicke und bis zu welchem Komplexitätsgrad erkennbar sind. Mit einem Betatron (JME X-ray Betatron 7,5 MeV) und einem Matrixdetektor (Perkin Elmer XRD 1621) wurden Aufnahmen von unterschiedlichen leichten Objekten angefertigt, die zunehmend in eine Umgebung aus Schwermetallen gestellt wurden. Mit unterschiedlichen Energieeinstellungen wurde untersucht, inwieweit eine Materialerkennung hinter welcher Abschirmung möglich ist. Die experimentellen Ergebnisse werden mit Simulationen verglichen, die mit einer Software zur Modellierung von Durchstrahlungsverfahren (aRTist) erzeugt wurden. Dabei wird dem Problem der Aufhärtung bei Anwendung von Bremsstrahlung Rechnung getragen. Die Ergebnisse können sowohl zur Erkennung von Fremdkörpern in Maschinen oder Pumpen als auch zur Überprüfung von Frachtladungen im Bereich der öffentlichen Sicherheit und bei Zollkontrollen dienen.
Statistic Evaluation of Sample Test Results to Determine Residual Strength of Composite Gas Cylinders
Authors: Mair Georg W. , Scherer Florian
Published: December 2013

Hydraulic pressure testing of composite gas cylinders damages the cylinders, but provides very limited information. For this reason, the Federal Institute of Materials Research and Testing (BAM) demands the assessment of the safety of design types for retest periods of more than three years by employing destructive testing of small samples. This gives insight in properties not quantified by current standards. Depending on the cycle fatigue behaviour of the individual design type either cycle tests or slow burst tests are employed. In this paper methods are introduced for a statistical assessment of sample results of both test methods. This includes gathering the average and load cycle burst strength, and the scatter of a sample as well as analysing and plotting them as a pair of values. Each pair of values representing a sample can be assessed approximately regarding its survival rate in service, if the introduced performance charts are employed.
NiCr Thin Film Strain Gauges Fabricated on Glass Substrates
Authors: Danışman Murat , Cansever Nurhan
Published: December 2013

In order to investigate the strain gauge characteristics of NiCr thin films, 500 nm NiCr (80 wt.-% and 20 wt.-%, respectively) thin films were deposited on glass substrates by DC magnetron sputtering. After deposition, NiCr thin films were characterized by using X-Ray diffraction analysis, scanning electron microscope and four-point probe techniques inview of crystallization, phases, film structure and electrical resistivity. After characterization, NiCr thin films were shaped into strain gauges by photo lithography and wet etching techniques. Strain gauges were tested with different loads, and strain values were calculated by comparing the results with commercial NiCr strain gauges with the same surface area. Resistivity change vs. strain was plotted, and the gauge factor of fabricated thin film strain gauges was evaluated as 1.23.
A New Design of Dental Implant Coating Using Functionally Graded Material
Authors: Hedia Hassan S. , Fouda Noha
Published: December 2013

Dental implantation treatment has developed into one of the most successful prosthetic technologies. A critical progress made in this area was the development of biocompatible materials to enable an engineered device (implant) to integrate within its surrounding bony tissues. Titanium and its alloys have been widely adopted as such materials due to their excellent biocompatibility. However, their mechanical properties largely differ from those in host bony tissues, which is problematical in osseointegration and bone remodeling. The challenge to face in prosthetics is to develop both biologically and mechanically compatible biomaterials for this purpose. Few existing research has been reported to develop an optimized design of functionally graded material (FGM) dental implant for promoting a long-term success. One of the authors of the present Contributions has previously designed a new FGM dental implant coating graded in axial direction from titanium at the apex to collagen at the basis of the dental implant. The aim of this investigation is to design a new gradation direction of FGM dental implant coating as well as studying the effect of coating thickness on the maximum von Mises stresses in bone adjacent to the coating layer. The gradation of the elastic modulus is changed along the longitudinal direction. Stress analysis using a finite element method showed that using a coating of 150 μm thickness, which is functionally graded from titanium at the outer shell adjacent to the bone to collagen at the inner shell adjacent to the implant, will reduce the maximum von Mises stress by 16 % and 13 % compared with the common conventional coating materials such as collagen and hydroxyapatite coatings, respectively. However, using FGM coating graded from hydroxyapatite at the outer shell to titanium at the inner shell reduces the maximum von Mises stress by 8 % and 5 % compared with collagen and hydroxyapatite coatings, respectively, but this gradation can improve the biocompatibility and can also achieve a full integration of the implant within the living bone, which increases the life of the implant.
The impact of non-optimized operation of nuclear power plants on the nuclear fuel burn-up
Authors: de Oliveira H. R. V. , Martinez A. S.
Published: December 2013

The purpose of this paper is to model the additional nuclear power and fuel burn-up needed to compensate non-optimized operation. Technical limitations, such as changes in reactivity, the additional nuclear power needed to compensate deficiencies of the electrical system control, and other related nuclear parameters are evaluated. The equations that define and quantify the additional nuclear power that must be generated by a reactor in order to fulfill the commitment with the national electric system operator are modeled in a simple and precise way.
Comparison of two modern approaches for MBDA calculations of RBMK-type reactors
Authors: Urbonas R. , Kaliatka A. , Ušpuras E.
Published: December 2013

The paper presents the comparative study of modern approaches, i. e. “best-estimate” and “partially-conservative” for the accident analyses based on calculations of Maximum Design Basis Accident at the Ignalina Nuclear Power Plant. The calculated critical parameters, values of which lead to the determination of the safety margin, were compared for both analyses. The comparison of the results shows that these values for “best-estimate” calculation are slightly lower in comparison with “partially-conservative” calculation results. That enables to draw a conclusion that in most cases both approaches – either “best-estimate” or “partially-conservative” can be applied. The latter approach looks tempting since in this case only one calculation is sufficient; while in the case of the “best-estimate” approach a high number of calculations is required. Thus, the “partially-conservative” approach requires considerably less computational time. However, when using this methodology the obtained calculated results do not meet the acceptance criteria, the complete analysis by employing “best-estimate” approach is necessary.
Monte Carlo simulation of the ETRR-2 research reactor using the MCNP Code
Authors: Aziz M.
Published: December 2013

The MCNP computer Code is used to model the ETRR-2 research reactor. A computer program was designed to evaluate the axial burn-up of the fuel elements. The excess reactivity of the reactor core is calculated for different core configurations and compared with the existing measurements. The thermal flux is also calculated and compared with measurements. Several factors that affect the safety of the reactor such as power peak and the effect of control rod insertion on the reactor power and flux were studied and analysed. The agreement between the MCNP results and the experimentally determined values is good.
Preliminary evaluation of the fixed and fluidized bed nuclear reactor concept using the IAEA-INPRO methodology
Authors: Sefidvash F.
Published: December 2013

The International Atomic Energy Agency (IAEA) through its INPRO Project has developed a methodology to evaluate the innovative nuclear reactors. Here is a preliminary application of this methodology to evaluate the Fixed and Fluidized Bed Nuclear Reactor Concept (FBNR). The FBNR is based on pressurized light water reactor technology. It is a small, modular, and integrated primary circuit reactor. It is an inherently safe and passively cooled reactor concept. FBNR in its first phase of development is a fixed bed nuclear reactor and in its next phase will be a fluidized bed nuclear reactor. FBNR in its advanced versions can use supercritical steam or helium gas as coolant, and utilize MOX or thorium fuel.
Analysis of a total loss of pool water accident in MTR-type research reactors
Authors: Yilmazer A. , Yavuz H.
Published: December 2013

In this study, the transient in which the pool water is lost throughout one or more of the main coolant pipes which are supposed to be broken guillotine-like is investigated for the TR-2 research reactor in Istanbul. The applicability of the methods used for other similar types of research reactors is shown. Decrease of the pool water level until the top of the core, and from the top to the bottom of the core are examined as two successive phases of the accident. Finite difference scheme and integral methods are employed to solve energy equations and the results of both methods are compared. The finite difference solution uses an explicit form for the analysis of the first phase, and a moving boundary approach for the second phase. The integral method is based on the assumption that the temperatures appearing in the energy equations have the same profiles during the transient as the steady state ones. Analyses are done both for nominal and hot channel, and the results of both methods are observed to be in agreement.
Air pollution forecasts of the German Weather Service for IMIS
Authors: Fay B. , Glaab H. , Jacobsen I. , Klein A.
Published: December 2013

The emergency response system of the German Weather Service (Deutscher Wetterdienst – DWD) will be described including the models (trajectories, LPDM), the national (IMIS, RODOS) and international (IAEA, WMO) co-operation. Examples show the performance of the models and the system.
Natural radionuclides in mineral water in Germany
Authors: Obrikat D. , Beyermann M. , Bünger Th. , Viertel H.
Published: December 2013

The activity concentrations of Ra-226, Ra-228, Po-210, Pb-210, U-234, U-235, U-238 and Ac-227 have been determined in approx. 400 bottled mineral waters in Germany. The median value of radiation exposure for children (age 0–1 year, 1701/y) is 0.047 mSv/a. The main part of exposure (80–90% approx.) is caused by the isotopes Ra-226 and Ra-228. The median values of concentrations are 7 mBq/l, respectively. Maximal concentrations up to 370 mBq/l were observed.
Strategy for taking measurements using the German Integrated Measuring and Information System (IMIS) in the case of a nuclear emergency
Authors: Bieringer J.
Published: December 2013

In the case of a nuclear emergency situation, as compared to normal operation, other requirements such as type of information, availability in time and amount must be met in order to characterise the radiological situation and provide recommendations for countermeasures. These requirements are dependent upon the different phases of the emergency. In the beginning phases (Phase 1 and Phase 2) the timely availability of information is of great importance. The focus changes more and more to the diversity of the analysed environmental media in the course of the event. The status at the plant, source term estimates, meteorological data, and weather forecasts are needed in order to calculate trajectories and assessing the prognoses of the expected external dose and contamination during Phase 1 before the release. In addition, on-line dose rate measurements and information on the radionuclide spectrum are necessary. The aim is to get an overview of the actual radiological situation and its possible development in time in order to introduce countermeasures, where necessary. In Phase 2, when the dispersion of radioactivity is finished, measurements according to § 3 of the Precautionary Radiation Protection Act (StrVG) are of primary interest, particularly the determination of contamination to human food and animal feed. During the phase following the deposition, the external dose as a function of time can be well predicted if the nuclide spectrum is known. In Phase 3, i. e. weeks following the emergency, measurements are used to observe the activity concentrations in different media over a longer period of time.
Theoretical studies on thermal stratification in a side heated cavity
Authors: Kumar N. V. Satish , Maheshwari N. K. , Vijayan P. K. , Saha D. , Sinha R. K.
Published: December 2013

The Advanced Heavy Water Reactor (AHWR) being designed in India is a vertical pressure tube type boiling water-cooled and heavy water moderated reactor. A key feature of the AHWR is the use of simple passive systems to respond to any postulated accidental condition. Passive Containment Cooling System (PCCS), is one such system, which provides long term cooling for the reactor containment following a Loss Of Coolant Accident (LOCA). The system consists of Immersed Condensers immersed in a gravity driven water pool. One of the important phenomena related to PCCS functioning is the effect of thermal stratification in the water pool. The heat transfer from IC to water pool takes place by natural convection. In this process, the pool water can get stratified. Towards the top of the water pool, the temperature of water may reach almost the boiling point while at the bottom of the pool, the water is likely to remain at the initial temperature. Hence, in order to study the integral system response of PCCS, the phenomenon of stratification requires to be investigated. Towards this objective, the case of a rectangular enclosure heated symmetrically on both the vertical sides was considered and a two-dimensional analysis was done by numerically solving the mass, momentum and energy equations. Theoretical results were generated by the use of a computer code developed for this purpose and the results are then compared with the experimental results available in literature.
Comments on the question of the discovery of element 112 as early as 1971
Authors: Brandt R.
Published: December 2013

There are two independent claims for the discovery of element 112: The claim by Hofmann et al. from 1996 and the older claim from 1971 by Marinov et al. This Comment will not challenge the experimental results of Hofmann et al., but it will discuss one aspect of the claimed discovery of element 112 by Marinov et al., as their experiment has never been reproduced in exactly the form in which the original experiment has been carried out. The reasons for this deficiency may not be found in the field of science, but possibly in radioprotection restrictions for researchers who want to carry out such an experiment. However, such is not a sufficient reason to exclude the original claim from all considerations of the responsible international authorities, who have to settle such priority questions. It may be in agreement with scientific traditions, that when the responsible international committees do not feel to be able to come to a positive decision on the “1971” claim, they could keep the priority problem unsettled for the time being.
Analysis of future nuclear power plants competitive investment costs with stochastic methods
Authors: Feretic D. , Cavlina N. , Grgic D.
Published: December 2013

One of main issues in planning future power system expansion is forecast of expected electricity generating costs. Due to the fact that the forecast should be extended to plants lifetime it is best to consider and compare lifetime levelized cost of generated electricity for each of the candidate plants. The analysis is a part of studies for the future expansion of the Croatian power system, but is valid generally. The main competitors to nuclear power plants are combined cycle natural gas fired plants whose economy strongly depends upon gas cost. The expected increase of gas cost in following decades can substantially influence the competitiveness of nuclear plants versus gas fired plants. In addition, the costs of electricity produced by coal fired plants and wind electricity generators were also considered. Due to uncertainties of input parameters it is advantageous to use probabilistic instead of deterministic method of analysis. The paper considers the problem of the probabilistic distribution of nuclear power plant investment costs for which such plants could be competitive to gas fired plants for an estimated probable range of gas cost increase rates.
Effects of spectral shifting in an inertial confinement fusion system
Authors: Şahin S. , Şahin H. M. , Yildiz K. , Acir A.
Published: December 2013

The main objective is to study the effects of spectral shifting in an inertial confinement system for kT/shot energy regime on the breeding performance for tritium and for high quality fissile fuel. A protective liquid droplet jet zone of 2 m thickness is used as coolant, energy carrier, and breeder. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF4 or UF4) starting by 2 moles% up to 12 moles%. Spectrum softening within the inertial confinement system reduces the tritium production ratio (TBR) in the protective coolant to a lower level than unity. However, additional tritium production in the 6Li2DT zone of the system increases TBR to values above unity and allows a continuous operation of the power plant with a self-sustained fusion fuel supply. By modest fusion fuel burn efficiencies (40 to 60 %) and with a few mol.% of heavy metal salt in the coolant in form of ThF4 or % UF4, a satisfactory TBR of > 1.05 can be realized. In addition to that, excess fissile fuel of extremely high isotopic purity with a rate of ∼ 1000 kg/year of 233U or 239Pu can be produced. Radiation damage through atomic displacements and helium gas production after a plant operation period of 30 years is very low, namely dpa < 1 and He < 2 ppm, respectively.
Simulation of the hydrogen distribution in the confinement of the Ignalina nuclear power plant
Authors: Babilas E. , Urbonavicius E. , Rimkevicius S.
Published: December 2013

The Accident Localisation System (ALS) of the Ignalina nuclear power plant is a “pressure suppression” type confinement, which protects the population, workers and environment from radiation hazards. According to the safety justification of RBMK-1500 reactors, ∼110 m3 of hydrogen is released to the ALS compartments during a maximum design basis accident. To prevent the possible hydrogen accumulation and combustion, a hydrogen concentration control system is installed. The main goal of the present paper is to present the ALS model for the COCOSYS code and assess the effectiveness of the hydrogen control system of the Ignalina nuclear power plant. The analysis performed identified locations of possible hydrogen accumulation in the ALS compartments during the transient processes and assessed the mixture combustibility in these locations for a maximum design basis accident scenario.
ICDE-results on complete common cause failures in the light of results obtained with the POS model
Authors: Berg H. P. , Fröhmel T. , Görtz R. , Kesten J. , Weil L.
Published: December 2013

Since April 1998, the OECD Nuclear Energy Agency (NEA) is operating the International Common Cause Data Exchange (ICDE) project. The objectives of this exchange comprise, to collect and analyze CCF events, to generate insights into root causes and derive preventing approaches, to strengthen experience feedback and, last but not least, to facilitate quantification. For a large number of about 1000ICDE events the fractions of complete CCFs have been displayed as a function of the degree of redundancy. Complete CCFs are relevant to PSA as in these cases the respective function fails and in many cases significant contributions to core damage can be expected. From the point of view of quantitative CCF modeling, the theoretical theoretical approach to of the empirically available complete CCF fractions represents a major challenge. The large number of events have been collected in the ICDE project for nine different types of components in nuclear power plants from ten different countries. Obviously, these components are operated under different conditions and technical specifications. Therefore, an approach is chosen which is based on low information distributions of the model parameters to account for the large variation throughout the component populations. The Process Oriented Simulation Model (POS) has been selected to carry out an analysis of complete CCFs. The consistency between the theoretical approach – considering its simplicity – and the data observed is satisfactory.
Analytical study of the closure flow inside the ETRR-2 core chimney
Authors: El-Morshdy S. El-Din
Published: December 2013

The present work is carried out in order to study the closure flow inside the core chimney of the Egypt Second Research Reactor (ETRR-2). Based on the finite difference technique, a two dimensional model is developed to simulate the coolant flow inside the chimney. The model is verified by FEHT finite element program. Then a study of different closure flow values inside the chimney was made using the developed model where a flow map is plotted showing the stagnation depth for each closure flow. The flow map shows that for a closure flow greater than 0.16 m3/h, no active water ascends from the core to the pool through the chimney. The model results are analyzed and discussed.
Incineration of weapon grade plutoniumin a (DT) fusion driven hybrid reactor using various coolants
Authors: Übeyli M. , Acır A.
Published: December 2013

Reducing separated weapon grade (WG) plutonium is very crucial since this plutonium may be misused and/or released accidentally into the environment. Burning WG plutonium in nuclear reactors can help to solve this serious problem effectively. In this study, the incineration of WG plutonium mixed with thorium in a (DT) fusion driven high power density hybrid reactor was investigated for an operation period of 24 months. Furthermore, the effect of using different coolants on the burning of WG plutonium in the hybrid blanket was examined. Numerical calculations showed that WG plutonium can be denatured to reactor grade level within a very short time by using a high power density hybrid reactor. Moreover, addition of WG plutonium to thorium improved the neutronic performance of the reactor significantly.
Calculation of the pin power distribution for a thorium reactor assembly and benchmarking
Authors: Aziz M.
Published: December 2013

A computer model was developed to perform neutronic and burn-up analysis for an assembly of a thorium reactor. The MCNP computer code was used to model the geometry of the assembly and to determine both, the power and flux distribution. A system of ordinary differential equations which represents all fuel isotopes was solved numerically to evaluate the time behavior of fuel composition and burn-up. The results of the present model were compared with the solutions of benchmark problems and satisfactory agreement was found.
Thermal-hydraulic modeling of reactivity accidents in MTR reactors
Authors: Khater H. , Abu-El-Maty T. , El-Morshdy S. El-Din
Published: December 2013

This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactor during a reactivity insertion accident (RIA). The model is formulated for coupling reactor kinetics with feedback reactivity and the reactor core thermal-hydraulics. Two channels are considered; the average and the hot channels to represent the reactor core. The developed computer program is compiled and executed on a personal computer using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. The model is used for simulating uncontrolled withdrawal of a control rod for the ETRR-2 reactor in transient with over power scram trip. The simulation results for the ETRR-2 are analyzed and discussed.
Prediction of the onset of flow instability in the ETRR-2 research reactor under loss of flow accident
Authors: Khater H. A. , El-Morshdy S. El-Din , Ibrahim M. M. A.
Published: December 2013

In the present work, prediction of the onset of flow instability (OFI) in Egypt's second Research Reactor (ETRR-2) under loss of flow accident (LOFA) conditions due to loss of off-site power has been performed based on the model developed in previous work. Calculations are performed for LOFA with Scram due to low flow signal at 90% of the nominal flow where the time length that covers this transient is 4.8 seconds. Both, the best-estimate and conservative calculations are performed under the specified operating conditions and geometrical data of ETRR-2 for both, exponential and ramp pressure gradient change. The OFI locus is predicted and plotted against the flow velocity, exit coolant temperature and bubble detachment parameter for several heat fluxes. The results are analyzed and assessed in order to obtain the safety margins for OFI phenomenon that measure how far the operating conditions are from the OFI locus. It is found that, the safety margins for OFI phenomenon in the best-estimate calculation are 2.62 and 2.35 for steady state and LOFA transient respectively, while its values in the conservative calculation are 1.68 and 1.47, respectively.
Application of the U

N
method to the reflected critical slab problem for one-speed neutrons with forward and backward scattering
Authors: Öztürk H. , Anlı F. , Güngör S.
Published: December 2013

The UN method is used to solve the critical slab problem for reflecting boundary conditions in one-speed neutron transport theory. The isotropic scattering kernel with the combination of forward and backward scattering is chosen for the neutrons in a uniform finite slab. It is shown that the method converges rapidly with easily executable equations. The presented numerical results are compared with the results available in the literature.
The OECD FIRE database
Authors: Angner A. , Berg H. P. , Röwekamp M. , Werner W. , Gauvain J.
Published: December 2013

Realistic modelling of fire scenarios is still difficult due to the scarcity of reliable data needed for deterministic and probabilistic fire safety analysis. Therefore, it has been recognized as highly important to establish a fire event database on an international level. In consequence, several member countries of the Nuclear Energy Agency of the OECD have decided in 2000 to establish the International Fire Data Exchange Project (OECD FIRE) to encourage multilateral co-operation in the collection and analysis of data related to fire events at nuclear power plants. This paper presents the OECD FIRE project objectives, work scope and current status of the OECD FIRE database after 3 years of operation as well as first preliminary statistical insights gained from the collected data.
Characterization of dose rate instruments for environmental radiation monitoring
Authors: Wissmann F. , Rupp A. , Stöhlker U.
Published: December 2013

Any dose rate instrument used to monitor environmental radiation must be characterized with respect to its response to secondary cosmic radiation and terrestrial radiation. This requires calibrations using different photon radiation fields, the measurement of the instruments internal background and the exposure to secondary cosmic radiation. The methods developed within the cooperation between BfS and PTB are explained in detail by an intercomparison of instruments which are part of the German early warning network IMIS and a high-pressure ionization chamber which was used as the reference instrument. The measurements were performed at the PTB facilities for the dosimetry of environmental radiation. It is shown how the ambient dose equivalent rate of the terrestrial radiation is determined and how the sensitivity of an instrument to a radioactive cloud passing by can be investigated.
Axial flux harmonics affecting the parallel channel instabilities in boiling water reactors
Authors: Aktaş D. , Geçkinli and M.
Published: December 2013

The existence of boiling phenomena in boiling water-reactors makes these systems susceptible to instabilities. Neutronics coupled with the thermal-hydraulics complicate the problem further. Boiling water reactor instabilities are classified as global or local according to the spatial dependency of the power fluctuations. At present, numerical codes can simulate these instabilities rather well on event basis; but more general physical insight can be gained readily through analytical modeling. One of the earliest models applied to the problem is the parallel channel instability, so called because of the applied thermohydraulic boundary conditions. In this work a new type of instability model is proposed and studied by including the effect of the first axial harmonic of the flux. Thermal-hydraulics is represented with one-dimensional two-phase homogeneous equilibrium flow model equations. Analysis is performed with linearized equations in the frequency domain. Special emphasis is given to the motion of the boiling boundary. The results clearly indicate that the first axial harmonic of the flux affects the stability of the reactor adversely, and must be accounted for to make more realistic predictions.
Uncontrolled withdrawal of a control rod without SCRAM
Authors: Abou-El-Maaty T.
Published: December 2013

In the present work, the thermal-hydraulic analysis of reactivity-induced transients in a Low Enriched Uranium (LEU) core of a typical material test research reactor is conducted using the previous program developed Khater et al. The analysis was done for the uncontrolled withdrawal of a control rod under scram-disabled conditions. The initiating event reactivity was considered with and without influence of the reactivity efficiency curve (“S” curve). The results of the transient calculations are analyzed and compared with each other. In the transient without the “S” curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down is occurring after 1.85 s. In the transient with the “S” curve influence, a super prompt highly critical situation is produced (1.762 $ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. A fast clad melt down is resulting in the hot channel at 1.088 s and a stable film boiling is occurring. This study shows that the influence of the reactivity efficiency curve results in higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down than that of a linear assumption.
Unified treatment of the P
(λ)

n
approximation to solve the reflected slab criticality problem with strong anisotropy
Authors: Yılmazer A. , Tombakoglu M.
Published: December 2013

Ultraspherical polynomial P
(λ)

n
approximations are used in reflected slab criticality calculations for strongly anisotropic scattering. A unique formulation is developed for reflected slab criticality conditions whose special cases are spherical harmonics approximations, Chebyshev polynomial approximations of first and second kind. It is shown in the calculations that all ultraspherical polynomial approximations are equivalent regarding the convergence of the solution of one-speed neutron transport equation for various degrees of anisotropy and cross section parameters. Although some ultraspherical polynomials approximations converge in the mean as the order of approximation is increased and some converge uniformly, they all converge to the same critical thickness which implies their equi-convergence. It is also demonstrated that the first order approximation P
(λ)
1 would give the exact critical thickness for a certain value of the parameter λ.
Application of theUNapproximation to the neutron transport equation in slab geometry
Authors: Bülbül A. , Ulutas M. , Anli F.
Published: December 2013

The critical slab problem of the one-speed and one dimensional neutron transport equation for an isotropic homogeneous medium was studied by using the Chebyshev polynomial approximation, i. e. UNmethod. The results obtained by the UNmethod for various c values, using the two type boundary conditions which are known as Mark and Marshak boundary conditions, are presented in tables. The tables also include the results obtained by the PNmethod for comparison.
Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
Authors: Horhoianu G. , Patrulescu I.
Published: December 2013

Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. A fissile content in the RU of 0.9 to 1.0 % makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. The Institute for Nuclear Research (INR) Pitesti has analyzed the feasibility of using RU fuel with 0.9 – 1.1 w% 235U in the CANDU-6 reactors of the Cernavoda Nuclear Power Plant (Cernavoda NPP). Using RU fuel would produce a significant increase in the fuel discharge burnup, from 170 MWh/kgU currently achieves with natural-uranium (NU) fuel to about 355 MWh/kgU. This would lead to reduced fuel-cycle cost and a large reduction in spent-fuel volume per full-power-year of operation. The RU fuel bundle design with recovered uranium fuel, known as RU-43, is being developed by the INR Pitesti and is now at the stage of final design verification. Early work has been concentrated on RU-43 fuel bundle design optimization, safety and reactor physics assessment. The changes in fuel element and fuel bundle design contribute to the many advantages offered by the RU-43 bundle. Verification of the design of the RU-43 fuel bundle is performed in a way that shows that design criteria are met, and is mostly covered by proof tests such as flow and irradiation tests. The most relevant calculations performed on this fuel bundle design version are presented. Also, the stages of an experimental program aiming to verify the operating performance are briefly described in this paper.
The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
Authors: Seifritz W.
Published: December 2013

The fundamental mode of a CANDLE burn-up wave, propagating into fertile material, in one-dimensional geometry was derived analytically with the result that the fluence and the neutron flux shapes turned out to be a Gudermannian and a Secans Hyperbolicus, respectively, of a Galilean variable. Furthermore, the fundamental evolution equation for the neutron flux was represented and discussed thoroughly. All other higher dimensional cases, however, have to be treated by numerical methods.
The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
Authors: Segatto C. F. , Vilhena M. T. , Marona D. V.
Published: December 2013

In this work we present a LTSN formulation for the isotropic neutron transport problem in a slab assuming c = 1. The LTSN solution for c = 1 is modified by applying the Schur decomposition and Heaviside expansion techniques. Numerical results are reported.
Optimization of the selection of analysis methods for the determination of naturally occurring radionuclides
Authors: Heinrich T. , Funke L. , Köhler M. , Schkade U.-K. , Ullrich F. , Löbner W. , Höpner J. , Weiß D.
Published: December 2013

The determination of naturally occurring radionuclides in the environment is a complex problem. Particularly the optimization of the analysis methods is very important, while the choice of the method influences the financial and personnel resources, as well as the quality of the results. This work is prepared from a working group within the working group(suggest instead use Research Group) “natural radioactivity” of the German-Swiss radiation protection association (FS). It is a recommendation for clients of existing laboratories, as well as new or renovated laboratories, which order measurements. This article shows some excerpts of the full work.
A study on the neutronic performance of the ARIES-RS fusion reactor with various coolants bearing nuclear fuel
Authors: Übeyli M.
Published: December 2013

Heavy metal molten salts have a good potential to be used in fusion reactors due to their capability of increasing fusion power and producing fissile fuel. In this study, the effect of various molten salts bearing nuclear fuel (99.5 % Flibe + 0.5 % Weapon grade (WG) PuF4, 99 % Flibe + 1 % WG PuF4, 99.5 % Flibe + 0.5 % Reactor Grade (RG) PuF4, 99 % Flibe + 1 % RG PuF4, 75 % LiF + 25 %ThF4, 75 % LiF + 24 % ThF4 + 1 % 233UF4 and 75 % LiF + 23 %ThF4 + 2 % 233UF4) on the neutronic performance of a magnetic fusion reactor called ARIES-RS was investigated. Neutron transport calculations were performed with the help of the code, Scale4.3. Numerical results showed that a significant enhancement on the neutronic performance was achieved with respect to the energy multiplication and fissile fuel breeding by keeping the sufficient tritium for the (DT) driver compared to the pure ARIES-RS fusion reactor for all investigated coolants except for the salts of 75 % LiF + 25 % ThF4 and 75 % LiF + 24 % ThF4 + 1 % 233UF4. Since these two salts did not satisfy tritium self-sufficiency.
Prediction of the time-dependent failure rate for normally operating components taking into account the operational history
Authors: Vrbanić I. , Šimić Z. , Šljivac D.
Published: December 2013

The prediction of the time-dependent failure rate has been studied, taking into account the operational history of a component used in applications such as system modeling in a probabilistic safety analysis in order to evaluate the impact of equipment aging and maintenance strategies on the risk measures considered. We have selected a time-dependent model for the failure rate which is based on the Weibull distribution and the principles of proportional age reduction by equipment overhauls. Estimation of the parameters that determine the failure rate is considered, including the definition of the operational history model and likelihood function for the Bayesian analysis of parameters for normally operating repairable components. The operational history is provided as a time axis with defined times of overhauls and failures. An example for demonstration is described with prediction of the future behavior for seven different operational histories.
Simulation model of a nuclear power plant turbine
Authors: Dutta A. , Goyal P. , Singh R. K. , Ghosh A. K.
Published: December 2013

A computer code “TURDYN” has been developed for prediction of high pressure and low pressure turbine torque under thermodynamic transient conditions. The model is based on the conservation laws of mass and energy. All the important components of turbine systems, e. g. high pressure turbine, low pressure turbine, feed heaters, reheater, moisture separator have been considered. The dynamic equations are solved simultaneously to obtain the stage pressure at various load conditions. The details of the mathematical formulation of the model and open loop responses for specific disturbances are presented.
Development and validation of the pressure surge computer code DYVRO mod. 3
Authors: Neuhaus T. , Schaffrath A. , Altstadt E.
Published: December 2013

Pressure fluctuations are generated in pipeline systems, when a fluid is rapidly accelerated or decelerated by fast closing or opening a valve, pump trip or start, breaking of pipes etc. TUEV NORD SysTec GmbH & Co. KG performs calculations of pressure surges in power plants and especially in nuclear power plants for many years. For this reason TUEV NORD has developed and qualified the pressure surge computer code DYVRO. This contribution describes in detail the actual code version DYVRO mod. 3 that has been modified with regard to the system of partial differential equations and the numerical scheme. Also the validation against representative experiments like the Simpson's experiment and an experiment at the Cold Water Hammer Test Facility CWHTF is shown. Results of DYVRO calculations are afterwards compared with results of own and external simulations with system codes (ATHLET and RELAP5) and with the pressure surge code WAHA. For this purpose the outcome of the EU project WAHALOADS that was supported within the 5th framework programme has been used.
A comparative study on classical polynomial approximations to the transport equation in spherical media albedo problems
Authors: Yılmazer A. , Kocar C.
Published: December 2013

The ultraspherical polynomial approximation which unifies all classical polynomial sequences in a unique form is used to calculate the albedo for isotropic scattering in a homogeneous spherical medium. This is the most general polynomial approach in the sense that it includes all classical polynomial methods to solve the transport equation such as and methods. For the first time an antisymmetric polynomial (ultraspherical polynomial) solution to the corresponding pseudo-slab problem is proposed. Very accurate and consistent albedo values are obtained for a variety of methods when compared to the literature. It is also shown that various approximations differ only in convergency characteristics; some converge monotonically, some in the mean.
The new IAEA safety standards on probabilistic safety assessment
Authors: Kuzmina I. , Lyubarskiy A.
Published: December 2013

Increasingly wide application of Probabilistic Safety Assessment (PSA) to support a variety of issues dealing with operation and design of Nuclear Power Plants (NPP) imposes certain requirements on PSA quality in terms of its scope, level of detail, approaches, assumptions, and data used. If a unified set of balanced requirements is developed and consistently implemented, it would give confidence in using PSA results and insights in different applications and Risk-Informed Decision Making (RIDM); in addition, it would be possible to consistently compare PSA results for different plants. Such requirements are currently developed or being developed in several countries. The International Atomic Energy Agency (IAEA) is in the process of development of several publications in the IAEA Safety Standards Series that would provide guidance on the standard technical content of PSA. The paper discusses the latest developments at the IAEA in the area of PSA and provides a brief overview of the publications.