bluuebunny/crossref_metadata_2025 huggingface dataset:
Wear Behaviour of Heat Treated 100Cr6 Steels
Authors: Polat Şeyda , Türedi Enbiya , Atapek Ş. Hakan , Köseoğlu Merve
Published: December 2013

In this study, several heat treatments were applied to DIN 100Cr6 steel to obtain different matrices. In the first stage of the study, solution annealing treatment was applied to the steel and cooling was carried out in various media (furnace, oil, and salt bath). In order to eliminate the stresses after transformation from austenization, a low temperature tempering treatment was applied to the quenched samples. All heat treated samples were examined using light microscopy after metallographic preparations. In the second step, ‘ball-on-disc’ type tribometer was used to determine the friction coefficient of the steels depending on the matrix phase. Weight loss was recorded and the friction coefficient versus distance was plotted for each steel. Worn surfaces of the steels were examined using scanning electron microscopy to characterize the wear mechanisms. It turned out that (i) pearlitic, bainitic and martensitic matrices could be obtained depending on the cooling medium, (ii) martensitic matrix had higher wear resistance based on its weight loss, (iii) abrasive and adhesive wear tracks were present on the worn surfaces of the steels.
Fatigue Behaviour Monitoring of an AISI 1045 Carbon Steel using the Statistical-Based Z-Notched Approach
Authors: Padzi Mahfodzah Md , Abdullah Shahrum , Nuawi Mohd Zaki
Published: December 2013

In this work, three percentages of constant loading are used for fatigue failure assessment. Fatigue behaviour under these three percentages of constant loadings is analyzed by using a new statistical-based approach known as integrated kurtosis-based algorithm for Z-notch filter (I-kaz). Fatigue tests were performed according to ASTM E466-96 standard with a strain gauge attached to the specimen being tested. A fatigue test was conducted under three loadings between 600–700 MPa and at a frequency of 8 Hz. A set of data acquisition systems was used to collect the fatigue strain signals. The I-kaz technique, which provides an I-kaz coefficient and a three-dimensional (3D) graphic, was used to describe and to evaluate the fatigue failure activities. The trends of I-kaz coefficients and of 3D graphics were compared to explain the fatigue damage induced by the three loadings.
Experimental and theoretical investigations on diffusion process for rare earth ores
Authors: He Ye , Li Wenzhi Z.
Published: December 2013

The diffusion reaction kinetics of weathered crust elution-deposited rare earth with mixed ammonium salts was studied. The influence of concentration of reagents and particle size of ore on diffusion rate was investigated. The results showed that the diffusion process and diffusion rate could be improved by increasing reagents concentration and decreasing diffusion flowing rate and particle size. The diffusion process could be explained with the shrinking core Model, which could be controlled by the diffusion rate of reacting reagents in porous solid layer.
Manual and Automatic Determination of the Plastic Strain Ratio
Authors: Bocz András , Márkus Dénes , Narancsik Zsolt
Published: December 2013

ISO 10113:2006 standard includes determination of plastic anisotropy (r-value). The r-value is one of the indices of the ductility of cold-rolled steel sheets. The article compares manual and automatic mathematical methods and their results.
Mechanical and Corrosion Resistance of (Ti,Cr)N Grown by DC Magnetron Sputter on H13 Tool Steel
Authors: Taweesup Kattareeya , Tungasmita Sukkaneste , Yongvanich Niti , Lothongkum Gobboon , Visuttipitukul Patama
Published: December 2013

This research studied properties of (Ti,Cr)N film prepared by co-deposition from dual unbalanced DC magnetron sputtering techniques at various temperatures. All samples were prepared at different growth temperatures for (Ti,Cr)N compound layer with atomic ratio Ti: Cr approximately 1: 1. The lattice parameter of the film is 4.18 Å, which is a value between those of TiN and CrN. It was found that both hardness and adhesion increased with an increase of growth temperature. The highest hardness value obtained is 24.78 GPa for film growth at 190 °C and film adhesion was significantly improved by increasing growth temperature. The growth temperature also plays an important role on corrosion resistance. Ecorr increases from −466 mV to −195 mV, while Icorr decreases from 188.21 nA/cm2 to 3.20 nA/cm2 with increasing the growth temperature to 190 °C. At highest growth temperature surface roughness decrease. The improved properties, both mechanical and corrosion properties, as well as surface uniformity are due to the change of film structure from columnar to equiaxial, which produces a denser film.
Messunsicherheit bei der Ermittlung des E-Moduls im Zugversuch an Stahl∗
Authors: Weißmüller Christian , Frenz Holger
Published: December 2013

Kurzfassung

Der E-Modul gehört zu den wichtigsten Werkstoffkennwerten. Die Ermittlung soll zukünftig in einem Anhang zur DIN EN ISO 6892-1 geregelt werden. Im Rahmen einer internationalen Eignungsprüfung wurden Richtigkeit und Messunsicherheit der E-Modulbestimmung untersucht. Die Bundesanstalt für Materialforschung und –prüfung stellte dazu besonders qualifizierte Referenzproben zur Verfügung. Die Ergebnisse zeigen, dass die Anforderungen an die einzusetzenden Messgeräte hoch sind und die berechneten Unsicherheitsbudgets zu fachlich sinnvollen Werten führen.
Investigation of Plasma Arc Cutting Parameters With Type-2 Fuzzy Set and System
Authors: Çelik Yahya Hışman , Özek Müzeyyen Bulut , Özek Cebeli
Published: December 2013

The objective of the present study was it to design a type-2 fuzzy set and system in order to predict surface roughness and hardness depending on the parameters (material thickness, cutting speed, arc voltage and current) of the plasma arc cutting process of S235JR sheet materials. Therefore, some experimental studies were conducted. The experimentally determined data were used to describe the type-2 fuzzy set and system. Type-2 fuzzy set and system was found to be usefull to predict surface roughnes and hardness. According to the obtained values, the best surface roughness and the values closest to the hardness of the raw material were obtained at 1500 mm/min cutting speed, 8 mm material thickness, 115 V arc voltage and 80 A current.
Current international activities to increase fire protection knowledge for nuclear power plants
Authors: Berg H. P. , Riese O. , Röwekamp M.
Published: December 2013

For nuclear power plants fire events are of high safety significance since they have the potential to damage more than one redundant safety train simultaneously. Within the frame of the recently started second series of periodic safety reviews, a probabilistic fire risk assessment has to be performed in addition to the deterministic analysis. Therefore, realistic fire occurrence frequencies as well as appropriate validated fire models and simulation codes are needed. All these complex issues require internationally harmonized activities to gain consistent, reliable and traceable information as well as to reduce the costs. Presently, three international activities are ongoing: the OECD-FIRE project with the goal to establish a quality assured and consistent international fire event database, and two further activities. In the frame of one of these fire models and computer codes shall be evaluated and validated based on real scale experiments under as far as possible realistic boundary conditions in order to support their extended use for decision making. The other one aims on analyzing fire induced electric failures of safety related systems and components. The objectives of the international co-operations and the first results are presented.
Accelerator driven systems for transmutation and energy production: challenges and dangers
Authors: Brandt R. , Birkholz W. , Shelaev I. A.
Published: December 2013

Accelerator Driven Systems (ADS) are an old technological idea: Relativistic proton accelerators deliver their beams onto massive heavy element targets, thus producing abundant neutron fluences. Placing this target into sub-critical nuclear fission assemblies is yielding substantial fission reactions, thus additional fission energy (Rubbia called such a system “Energy Amplifier”). This technology has recently attracted considerable attention due to advances in the construction of powerful accelerators. It allows the safe and cheap production of nuclear energy simultaneously with the destruction (Transmutation) of long lived radioactive waste, in particular plutonium and other minor actinides (neptunium and americium). The principles and the present-state-of-the-art are described, including first experiments to transmute plutonium this way. This technology needs, however, many more years of further “research and development” before large scale ADS's can be constructed. It may be even necessary to investigate the question, if all basic physics phenomena of this technology are already sufficiently well understood.
Calculation of control plates life time and worth at the ETRR-2 research reactor
Authors: Aziz M.
Published: December 2013

A detailed three-dimensional MCNP model of the ETRR-2 research reactor has been developed for the analysis of the neutronic parameters of the reactor. The model was used to determine the partial control plates reactivity worth and the behavior of the worth with time. The results of the present model were compared with the experimental measurements for partial control plates reactivity worth and with the design calculations for control plates behavior with time. The comparisons indicate satisfactory agreement.
Optimisation of occupational exposures during site remediation at WISMU
Authors: Schmidt P. , Wolf F.
Published: December 2013

Since 1991 the national corporation WISMUT GmbH has been charged with the rehabilitation of the sites of former Uranium mining and milling in East Germany. Since decommissioning and site remediation are carried out at places with enhanced dose rates and high concentrations of airborne radioactivity (Rn-222, long-lived alpha emitters), a large part of the employees is classified as occupational exposed radiation workers. Thus, in 2002, 1,073 workers were engaged with relevant works in radiation protection areas aboveground (radiation exposed workers of category B; mean effective dose: 0.3 mSv; maximum dose: 2.2 mSv), and 514 workers still prepared by underground activities the flooding of the Uranium mines (category A; mean effective dose: 0.9 mSv; maximum effective dose: 4.9 mSv). The paper presents the measures taken by WISMUT to optimise the occupational exposure. The need for implementing appropriate ALARA methods in the field of occupational exposure is last but not least caused by the fact that for a number of employees the integral exposure over the total working life time has exceeded already the 400 mSv figure.
Airborne measurement of radioactivity by learjet 35A
Authors: Dyck W. , Brust H. , Dalheimer A. , Steinkopff Th.
Published: December 2013

In case of a radioactive emergency radioactive plumes may be expected in heights up to the tropopause. For this reason aircraft measurements are necessary as a supplement to ground-based measuring sites. The authors present an aircraft type Learjet 35A used for localizing and analyzing radioactive contaminated air masses up to 12,000 m. The aircraft is equipped with an isokinetic high-volume sampler to collect aerosols on filters outside the pressurized cabin and an onboard gamma spectrometry system with a high purity germanium detector to analyze the filters during flight as well as a dose ratemeter, a radon daughter monitor and equipment to collect gaseous iodine and noble gases. The meteorological parameters temperature, atmospheric pressure, relative humidity, wind speed and wind direction are determined as well. Resulting data are transmitted from the plane to the headquarter of the German Meteorological Service (Deutscher Wetterdienst, DWD) and used for verifying the prognosis.
Survey of decommissioning projects
Authors: Rehs B. , Kaulard J.
Published: December 2013

Worldwide, 104 power and prototype reactors, 382 research reactors and 247 facilities of the nuclear fuel cycle have been shut down and are in different phases of decommissioning or decommissioning has already been completed. With 18 shut down power and prototype reactors, 11 shut down research reactors with a power of more than 50 kWth and with 10 shut down facilities of the nuclear fuel cycle Germany belongs to those countries with a lot of experience in decommissioning. This pertains not only to the number of shut down installations but also to the variety of different types. Further countries with a high degree of experience in decommissioning are among others the United States of America, the United Kingdom and France.
Decommissioning of gas-cooled reactors
Authors: Sterner H. , Cross M. T. , Rodriguez A.
Published: December 2013

Gas-cooled reactors represent an important part of the world's nuclear programme. A wide spectrum of designs has been developed and tested. The majority of these test reactors and 1st generation of power reactors has been shut down and is now in different stages of decommissioning. This is also the case for the pilot power reactors of the types High Temperature Gas-cooled Reactor (HTGR) and Heavy Water Gas-cooled Reactor types (HWGCR). Gas-cooled reactors show major generic design characteristics different from Light Water Reactors (LWR), hence influencing the choice of decommissioning strategy, waste/material management and also to a certain extent the decommissioning techniques and procedures applied. The most important issue influencing the decommissioning is the presence of large amounts of graphite in most of these reactors. The majority of the shut-down reactors is located in the UK and in France and major decommissioning programmes are under way in these countries. Interesting decommissioning work has also been carried out in Germany and in the USA. Most reactors are in a stage of safe enclosure. However, decisions have been taken recently to perform direct complete dismantling, notably in France, Italy and, to a certain extent, in Germany.
Decommissioning of reactors after accidents
Authors: Ackermann L.
Published: December 2013

The decommissioning is the final phase in the life cycle of a nuclear reactor. In case of finishing the operation after the foreseen lifetime the decommissioning is planned long-dated with regard to the options “immediate dismantling”, “safe enclosure” and “entombment”. In case of an accident the operational phase was mostly finished immediately depending from the possible options for managing the damaged facility. With regard to the accidents in the nuclear power plants Three Mile Island Unit 2 (USA), Lucens Experimental Reactor (Switzerland), A-1 (Slovakia) and Chernobyl-4 (Ukraine) this article presents a short description of these accidents and of the related decommissioning projects. The examples demonstrate that in principle these projects represent variations of the above mentioned options. It has, however, to be noted that the realisation of the entombment of the destroyed unit Chernobyl-4 is unique with regard to the essential high complexity related to the first and the second Sarcophagus.
Positive temperature reactivity coefficient of a TRIGA reactor at room temperature
Authors: Žagar T. , Ravnik M.
Published: December 2013

A positive isothermal temperature reactivity coefficient has been measured at zero power conditions at the TRIGA research reactor of the J. Stefan Institute in Ljubljana. The coefficient is measured in a critical reactor for different core configurations at room temperature and at low power. The reactor is filled with standard, 20 % enriched, fuel elements containing 12 wt % uranium with 6 % average burn-up. The experiments are analysed by reactor calculations using WIMSD code to provide the reactor physics explanations of the results measured. The positive temperature reactivity effect can be explained by reduced absorption in water prevailing at low fuel temperature over negative temperature effects on the reactivity in the fuel elements. The positive effect is due to the thermal spectrum shift in water and is not the consequence of a change in water density. The positive effect decreases with increasing temperature. The negative reactivity effect in fuel prevails at fuel temperatures above 50°C. The positive isothermal temperature reactivity coefficient has no effect on normal reactor operation in steady state and pulse mode.
Design stability and safety margins of the ETRR-2 core cooling system regarding seismic loads
Authors: Ahmed K.
Published: December 2013

The seismic design of a nuclear power plant includes two levels of design earthquake, the safe shutdown earthquake (SSE) and the operating basis earthquake (OBE). The OBE and SSE are considered in the nuclear power plant design as required by the IAEA safety regulations. A typical piping model for the ETRR-2 core cooling system includes ASME-class 1, 2 and 3 piping and was analyzed with regard to both IAEA safety regimes. A load combination, as stated by the IAEA Safety Guide 50-SG-S2, has been considered. The safety guide requires that the OBE should equal to at least one half SSE and that the plant should be shut down if exposed to earthquake intensity greater than the OBE. The results reflect the requirement for the precise design of a supporting system to accommodate for higher seismic peaks more than 0.2 g and the system needs to be more flexible.
Safety and security of radioactive sources – international provisions
Authors: Czarwinski R. , Weiss W.
Published: December 2013

For more than 50 years radioactive sources are used beneficially world-wide in medicine, industry, research and teaching. In the early 50ies mainly Ra-226 sources were used especially for medical applications. In the mean time a great number of radio-nuclides with more or less risk to individuals, society and environment are used. The number of these sources is increasing. The available experience with the application of sealed sources in industry, medicine, research and teaching shows that despite the widespread use of such sources a high level of safety can be achieved. One precondition is that the regulatory control of a radioactive source has to be carried out consistently during the life cycle of the sources – “from cradle to grave”. Particular attention has to be given to the so-called orphan sources which are not subject to regulatory control, either because they have never been under control, or because they have been lost, misplaced, abandoned, stolen or transferred without proper authorisation. The concern about orphan sources arising from poor safety and security standards of radioactive material around the world resulted in intensive global actions especially in the light of the security situation after the 11th September 2001. The improvement of regulatory control is one of the key elements in preventing people, goods and environment from being exposed exceptionally by the misuse of radioactive sources. Important steps toward the improvement of the safety and security of high radioactive sources are the IAEA Code of Conduct on the Safety and Security of Radioactive Sources and the European Directive on the Control of High Activity Sealed Radioactive Sources and Orphan Sources.
CCF analysis for new reactor designs
Authors: Blombach J. , Bordihn S. , Kollasko H.
Published: December 2013

Common Cause Failures (CCF) have been an issue of safety consideration in nuclear safety since a long time. For the design of the European Pressurised Reactor (EPR) probabilistic safety goals in conjunction with the limit set by CCF on failure probability that can be achieved by redundancy led to the requirement to fulfil vital safety functions by functionally diverse systems. The probabilistic safety analysis (PSA) which was performed for EPR takes into account CCFs which are modelled using the beta-factor model of the European Utility Requirements (EUR). The PSA shows that the probabilistic safety goal for core damage frequency – CDF < 10−5/year – is met with sufficient margin. In addition to the modelling of CCF in the PSA a deterministic CCF analysis is being performed. It is a safety function based approach on component level analysing frontline and backup systems with respect to common components or type identical components. It ensures that diverse safety functions are fulfilled by systems being sufficiently diverse including all necessary support systems. Components are identified where further diversification or additional defences against CCF are required.
CCF analysis in PSA applications from a licensee view
Authors: Schubert B.
Published: December 2013

In highly redundant systems CCF dominate the calculated unavailabilities. CCF contribute over 90% to the resulting core damage frequencies of some PRAs for German Nuclear Power Plants. Because the results of PRAs can lead to costly back fitting measures, it is essential to ensure these results. The principles of CCF analysis and the efforts to establish a comprehensive data base are described. In addition the utilities have undertaken efforts to develop concepts for defensive strategies and to introduce these measures in the plant operation in order to reduce the CCF probabilities. Unfortunately these efforts are yet not fully supported by the existing models.
Multigroup neutron reflection and leakage conditions for axial core models with anisotropic scattering
Authors: de Abreu M. P.
Published: December 2013

In this article, multigroup discrete ordinates interface conditions for neutron reflection and leakage were derived for axial models of nuclear reactors with anisotropic scattering. These conditions are completely free from spatial truncation error. The key to the derivation is a suitable reformulation of exact discretized equations – the multigroup spectral nodal equations – defined in the nonactive boundary regions at top and bottom. The resulting interface conditions can be used to replace the nonactive boundary regions of a nuclear reactor core in axial reactor computations. In order to illustrate the improvements that can be experienced by most one-dimensional multigroup discrete ordinates codes with our multigroup interface conditions, we perform a numerical experiment with a representative multigroup code and an axial core model with anisotropic scattering.
Analyses of severe accident scenarios in RBMK-1500
Authors: Kaliatka A. , Rimkevicius S. , Uspuras E. , Urbonavicius E.
Published: December 2013

Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena.
Comparative assessment of methods for the reactivity measurement in subcritical systems by pulsed experiments
Authors: Canepa S. , Dulla S. , Ravetto P.
Published: December 2013

The determination of the subcriticality level for source-driven systems is an important problem that can be handled by several deterministic methods developed for the interpretation of flux signals in pulsed experiments. The paper compares the performance of several methods for different physical characteristics of the system considered and for different durations of the source pulse. The effects of the subcriticality level and of the experimental uncertainty are also investigated. Furthermore, an analysis of the influence of spatial effects is carried out. The study gives some indications on the most suitable method in a specific experimental configuration pointing out at the situations where non-conservative evaluations of the reactivity can be expected, i.e. when the system effective multiplication constant is underestimated.
Quantitative application of Monte Carlo simulation in Fire-PSA
Authors: Keski-Rahkonen O. , Mangs J. , Hostikka S. , Korhonen T.
Published: December 2013

In a power plant a fire cell forms the basic subunit. Since the fire is initially located there, the full-scale time dependent fire simulation and estimation of target response must be performed within the fire cell. Conditional, time dependent damage probabilities in a fire cell can now be calculated for arbitrary targets (component or a subsystem) combining probabilistic (Monte Carlo) and deterministic simulation. For the latter a spectrum from simple correlations up to latest computational fluid dynamics models is available. Selection of the code is made according to the requirements from the target cell. Although calculations are numerically heavy, it is now economically possible and feasible to carry out quantitative fire-PSA for a complete plant iteratively with the main PSA. From real applications examples are shown on assessment of fire spread possibility in a relay room, and potential of fire spread on cables in a tunnel.
Source term assessment as a basis for protective measures for the population in case of a nuclear accident in a nuclear power plant with radiological consequences
Authors: Zander N. , Löffler H.
Published: December 2013

The reliability of the evaluation of the radiological situation in the pre-release phase of a nuclear accident strongly depends on the quality of the source term estimation. Input data for the decision support system RODOS, such as time, composition, extent, and duration of a possible release of radioactive substances (source term) serve as a basis for the assessment of possible radiological consequences in the vicinity of a plant. With the collaboration of BfS and GRS a source term estimation tool was tested at a reference plant in order to provide data for RODOS.
Fast online system for forecasting environmental impact during an incident
Authors: Kümmerle E. , Möllmann-Coers M. , Pomplun E. , Hille R.
Published: December 2013

In an incident, the operational control staff needs a rapid general view of the radiological situation in the vicinity of the emission source which is suitable for decision support. Because the measuring team cannot provide sufficient measurements immediately, model calculations based on the available information about the course of the accident are additionally required. Such calculations do not only serve to create an overview of the current situation but can also provide prognoses for several hours. For this purpose, a software system for environmental impact forecasting was developed at Research Centre Jülich, which calculates time-integrated doses as well as ambient gamma dose rates and activity values. The latter can be directly compared with values measured during the incident and consequently can be used to assess the model assumptions.
Harnessing monitoring measurements in urban environments for decision making after nuclear accidents
Authors: Kaiser J. C. , Pröhl G.
Published: December 2013

This article gives an overview on the conceptual design of the Inhabited Areas Monitoring Module IAMM which will be introduced into European decision support systems for nuclear emergencies. It will improve the use of monitoring data of radioactive contamination in urban environments for decision making. IAMM converts the dated gamma dose rate (GDR) measurements from geo-referenced locations into maps of surface contamination with an enhanced spatial resolution. Depending on the availability of the monitoring data, IAMM relies on two modes of operation. If there are only a few measurements, these are taken to improve the maps from a deposition model using data assimilation. If the number of measurements is sufficient to apply spatial interpolation IAMM will rely entirely on monitoring data. Suitable geo-referenced data points will be interpreted by IAMM with respect to their detector environment using the concept of location factors. The endpoints of IAMM can be used directly for decision making or dose calculations with either simple dose models or the more refined EuRopean Model for INhabited areas (ERMIN).
Protection against dependent failures, analysis of dependencies and derivation of CCF data
Authors: Knochenhauer M. , Mankamo T. , Hellström P. , Johansson G.
Published: December 2013

The article describes a number of activities that were part of a major Nordic co-operation project which was recently performed, and which aimed at providing state-of-the-art guidance on the handling of dependent failures in nuclear power plants, both in operation and safety management, and as a part of plant PSA:s. This includes a rather wide scope of activities ranging from reviewing and describing available sources of data on CCF and CCF modelling techniques to the analysis of the efficiency of various protective measures against dependent failures in safety related equipment. A set of guidance documents was issued. The project also included a number of data analysis pilot applications, where CCF events in the ICDE database were analysed applying the impact vector method for several types of safety critical components.
Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
Authors: Vaidya A. M. , Maheshwari N. K. , Vijayan P. K. , Saha D.
Published: December 2013

Three dimensional CFD simulations of the moderator flow in the calandria vessel of a heavy water reactor are performed using the PHOENICS CFD code. The model includes the entire calandria vessel consisting of three shells, calandria tubes and inlet and outlet nozzle openings. The computational model prepared in PHOENICS consists of (a) standard k-∊ turbulence model, (b) PARSOL technique for handling curved objects in cartesian grids and (c) Boussinesq formulation for handling variable density flows. PHOENICS is validated by applying it to three different flow cases. The flow pattern in the calandria vessel under normal operating conditions is obtained through simulation. The effect of the presence of calandria tubes and heat generation on moderator flow pattern is studied. The simulation is also performed for various heat loads and moderator mass flow rates. The maximum temperature achieved by the moderator flow under various heat loads and moderator mass flow rates is obtained.
Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
Authors: Babilas E. , Urbonavicius E. , Rimkevicius S. , Kaliatka A.
Published: December 2013

The Accident Localisation System (ALS) of the Ignalina nuclear power plant is a “pressure suppression” type confinement, which protects the population, employees and environment from radiation hazards. According to the emergency operating procedures for the Ignalina nuclear power plant ∼110 m3 of hydrogen is released to the ALS compartments during the maximum design basis accident. However, in case of beyond design basis accidents, when the oxidation of zirconium starts, the amount of generated hydrogen could be significantly higher. The volume concentration of hydrogen in the compartment might reach 4 % of a combustible mixture. To prevent a possible hydrogen accumulation in the ALS of the Ignalina nuclear power plant during an accident, the H2 control system is installed. The analysis performed resulted in the identification of places of possible H2 accumulation in the ALS compartments during transient processes and the assessment of the mixture combustibility in these places for a postulated accident scenario. The results also allow to evaluate the influence of the proposed severe accident management strategy on the ALS conditions.
Notifiable work activities in water supply facilities – execution experience from Bavaria
Authors: Reifenhäuser C. , Körner S.
Published: December 2013

Since 2001, the protection against natural radiation is implemented in the German Radiation Protection Ordinance. Water supply facilities are now obliged to estimate the radon exposure of their staff. In 75 water supply facilities in Bavaria, Germany, radon exposures above the action limit of 6 mSv per year were determined. In these facilities the radon exposure has to be continuously monitored. By the proposal of the Bavarian Environment Agency (LfU), this should be done by continuous wearing of personal track-etch detectors, that are changed every three months. The results are sent to the LfU and reported to the central German dose register. Because of measuring periods of three months a high radon exposure that could probably exceed the annual limit of 20 mSv can be recognised at an early stage and immediate protection measures can be taken. In collaboration with the University of Regensburg remedial strategies for the permanent reduction of the radon exposure have been developed. The exposure is effectively reduced by reducing the time spent inside the units as much as possible. A very effective way to reduce the radon concentration inside buildings and hence the radon exposure is to blow fresh air directly into the work place of the staff with stationary or mobile devices. An alternative way to reduce the indoor air concentration is to avoid any transfer of radon polluted air exhausted from purification tanks to indoor air. For a long lasting success of the remedial measures a work instruction was developed by the LfU. All plants, their mean radon concentration and the measures to be taken have to be stated there.
Radiological evaluation of mine water discharges from hard coal mining in the Fossa Eugeniana Area, Germany
Authors: Wanke C. , Ritzel S. , Sachse R. , Michel R.
Published: December 2013

Discharges of radium-carrying mine waters from hard coal mining in the area of the Fossa Eugeniana and the Rheinberger Altrhein in Germany have caused elevated radioactivity in sediments and soils. Extensive measurements of the ambient dose rate revealed some small area contaminations as well as a larger one in a nature protection area at the Rheinberger Altrhein. For the evaluation of the potential radiation exposure investigations of the activity concentrations of river sediments, soils, and biological materials were performed. The potential exposure of members of the public was modelled on the basis of the German Guidelines for the Calculation of Radiation Exposure due to Radioactivity from Mining Activities. In the assessment the present situation as well as its possible future development was considered.
Consideration of statistical uncertainties for the determination of representative values of the specific activity of wastes
Authors: Barthel R.
Published: December 2013

The German Radiation Protection Commission has recommended “Principles and Methods for the Consideration of Statistical Uncertainties for the Determination of Representative Values of the Specific Activity of NORM wastes” concerning the proof of compliance with supervision limits or dose standards according to § 97 and § 98 of the Radiation Protection Ordinance, respectively. The recommendation comprises a method ensuring the representativeness of estimates for the specific activity of NORM wastes, which also assures the required evidence for conformity with respect to supervision limits or dose standards, respectively. On the basis of a sampling survey, confidence limits for expectation values of specific activities are determined, which will be used to show that the supervision limit or the dose standard is met or exceeded with certainty, or that the performed sampling is not sufficient for the intended assessment. The sampling effort depends on the type and the width of the distribution of specific activities and is determined by the position of the confidence interval with respect to the supervision limit or of the resulting doses with respect to the dose standard. The statistical uncertainties that are described by confidence limits may be reduced by an optimised extension of the sample number, as far as necessary.
Assessment of FEMAXI and TESPA-ROD codes for modelling of BDBA in RBMK-1500
Authors: Jusevičiu-tė A. , Kaliatka A. , Urbonavičius E. , Duškesas G. , Juodis L. , Sonnenburg H. G.
Published: December 2013

Processes occurring during the reactor operation and accidents change the thermal, chemical and other properties in nuclear fuel rods. This paper describes the processes in the nuclear fuel rods and tools for simulation at these processes. Fuel rods used at Ignalina NPP were simulated with FEMAXI-V (Japan) and TESPA-ROD (Germany) codes. Developed models were employed for the analysis of processes in fuel rods in case of large LOCA Beyond Design Basis Accident (BDBA). The developed models and results of the analysis are presented in this paper.
An analytical solution for the time-dependent SN transport equation in a slab
Authors: Segatto C. F. , Vilhena M. T. , Gonçalez T. T.
Published: December 2013

In this work an analytical solution for the time-dependent SN transport equation in a slab using the double Laplace transform technique is reported. This procedure leads to an analytical solution for the angular flux in integral form. Numerical simulations are presented and compared with results in the literature.
Design Concept of the High Performance Light Water Reactor
Authors: Schulenburg T. , Bittermann D. , Starflinger J.
Published: December 2013

The “High Performance Light Water Reactor” (HPLWR) is a Light Water Reactor operating with supercritical water as coolant. At a pressure of 25 MPa in the core, water is heated up from 280 to 500 °C. For these conditions, the envisaged net plant efficiency is 43.5 %. The core design concept is based on a so-called “3-pass-core” in which the coolant is heated up in three subsequent steps. After each step, the coolant is mixed avoiding hot streaks possibly leading to unacceptable wall temperatures. The design of such a core comprises fuel assemblies containing 40 fuel rods and an inner and outer box for a better neutron moderation. Nine of these are assembled to a cluster with common head- and foot piece. The coolant is mixed inside an upper and inside a lower mixing chamber and leaves the reactor pressure vessel through a co-axial pipe, which protects the vessel wall against too high temperatures.
Performance analyses of the communication networks of a modern supervision and control system of research reactors
Authors: El-Madbouly E. I. , Shaat M. K. , Shokr A. M. , Elrefaei G. H.
Published: December 2013

The functions of the Instrumentation and Control (I&C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated.
Optimizing the dynamic response of the H. B. Robinson nuclear plant using multiobjective particle swarm optimization
Authors: Elsays M. A. , Aly M. Naguib , Badawi A. A.
Published: December 2013

In this paper, the Particle Swarm Optimization (PSO) algorithm is modified to deal with Multiobjective Optimization Problems (MOPs). A mathematical model for predicting the dynamic response of the H. B. Robinson nuclear power plant, which represents an Initial Value Problem (IVP) of Ordinary Differential Equations (ODEs), is solved using Runge-Kutta formula. The resulted data values are represented as a system of nonlinear algebraic equations by interpolation schemes for data fitting. This system of fitted nonlinear algebraic equations represents a nonlinear multiobjective optimization problem. A Multiobjective Particle Swarm Optimizer (MOPSO) which is based on the Pareto optimality concept is developed and applied to maximize the above mentioned problem. Results show that MOPSO efficiently cope with the problem and finds multiple Pareto optimal solutions.
Thermal analytical model for MoO3 ampoules irradiated in a high neutron flux
Authors: Abd-El-Hady A. , Abou-El-Maaty T.
Published: December 2013

A one dimensional steady state thermal analytical model has been developed to study the heat transfer and temperature distributions in a quartz ampoule filled with MoO3 powder. The source heat generation inside the ampoule is released from the high neutron flux (1.4 × 1014 neutrons/cm2s) interaction with MoO3. Natural and forced convections heat transfer boundary conditions are adopted during the irradiation process. The peak temperatures in MoO3 powder and quartz are calculated and compared with their melting temperatures to ensure the irradiation safety criteria.
Integral analysis of low power and shutdown operation: insights and outcome
Authors: Spitzer C.
Published: December 2013

Probabilistic Safety Assessments (PSAs) in Germany have essentially been conducted in the framework of recurrent Safety Reviews (SRs). Besides, probabilistic investigations have also been submitted and reviewed in the usual regulatory procedures. Significant safety relevant insights have e. g. been gained from the Human Reliability Assessment (HRA) in the context of Low Power and Shutdown (LPS) PSAs whose performance has been launched by the regulatory body. In this paper major insights and results together with the practical outcome particularly from Human Factor (HF) and safety relevance points of view are described. Thereby, general key issues on HRA as well as substantial insights and conclusions derived from LPS HRAs are addressed. Further, a specific example with respect to the design and review of a complete set of operating procedures specifically dedicated to the shutdown operation is illustrated by outlining the general approach as well as the overall outcome achieved. Finally, conclusions are drawn regarding the relevance of probabilistic investigations, the appropriateness of a symptom oriented proceeding related to shutdown operation and the potential of utilising an advanced HF methodology.
PSA Level 2 – An optimized methodology
Authors: Torri A.
Published: December 2013

A methodology for a Level 2 PSA is described that provides a more efficient structure than the two methods recommended in the German Leitfaden. This is accomplished by merging the top events for the system and accident management functions from the Level 2 space, referred to as the containment system event tree (CSET) as a direct extension of the Level 1 model and by introducing the containment phenomenological event tree (CPET) to represent only the phenomenological Level 2 top events. The interface between the Level 1+ CSET and the CPET is defined as the plant damage states (PDS) with the focused objective to minimize the effects of sequence to sequence variability on the CPET failure fractions. A clear and consequent distinction is made between the statistically random events modeled in the Level 1+ CSET and the deterministic physical processes with large state of knowledge uncertainties modeled in the CPET. It is shown that as a logical consequence the resulting CPET failure fractions are numbers and not distributions, and these numbers are consistent with the interpretation as the degree of confidence probabilities for the outcome of uncertain physical processes. While both methods recommended in the Leitfaden will yield correct results if properly applied, the proposed method will result in a more streamlined model that will be easier to trace, understand and review, and it will result in lower release category uncertainties by minimizing the sequence to sequence variability through the elimination of the intermediate pinch points and through the focused use of the PDS.
Analysis of external flooding and tsunamis for nuclear power plants at tidal rivers
Authors: Berg H.-P. , Winter C.
Published: December 2013

The German regulatory body has issued technical documents on methods and data for probabilistic safety assessment, elaborated for a comprehensive integrated safety review of all nuclear power plants in operation. These documents contain, among others, a graded approach for the probabilistic assessment of external flooding. Main aspects are explained such as the underlying probabilistic considerations and the mathematical procedures for the calculation of exceedance frequencies for nuclear power plants at tidal rivers in Germany. Morever, extreme events such as tsunami waves have been investigated.
The Hamburg polonium-210 case: Monitoring of internal contaminations in a malevolent act
Authors: Dalheimer A. , König K. , Noßke D. , Schäfer I.
Published: December 2013

The dead of Alexander Litvinenko in London in the late 2006 was linked with the visit of Dimitri Kovtun. So an investigation was started by the Hamburg Police. As the presence of polonium-210 was uncertain, the action was supported by the Federal Defense against Nuclear Hazards (NGA) of the Federal Office for Radiation Protection (BfS). Radiation measurements in some flats in Hamburg confirmed the suspicion of polonium-210 contaminations. As incorporations by ingestion or inhalation could not generally be excluded, the Coordinating Office on Incorporation Monitoring of the BfS (Leitstelle Inkorporationsüberwachung des BfS) was requested to clarify this issue. The BfS office then coordinated the required bioassay analyses of urine samples carried out at four officially appointed in-vitro laboratories for incorporation monitoring. The group of people to be examined included the members of the family involved living in the contaminated flats, the emergency forces, as well as potentially exposed members of the public. The calculation of the internal dose was based on standard assumptions and also conducted by the BfS. The internal radiation doses proved to be within the range of radiation doses due to natural exposures. Retrospectively it should be stressed that the collaboration and cooperation of all institutions involved in the operation in Hamburg was extremely good despite their differing organisational structures. As a result of the incident in Hamburg various middle and long-term measures have been taken since then by the BfS Coordinating Office on Incorporation Monitoring.
Assessment of radiological consequences due to routine atmospheric discharges from nuclear power plants
Authors: Kocar C. , Sökmen C. Niyazi
Published: December 2013

The radiological consequences of operational atmospheric discharges from nuclear power plants are analyzed by using coupled long-range atmospheric dispersion models with time-dependent dose and risk assessment models. The coupled model is applied to atmospheric discharges from twelve assumed nuclear power plants located at three different sites in Turkey. Individual as well as collective doses and risks are analyzed. The effects of multi-location food supply and interregional movements of people in the computational domain are also investigated while computing individual effective dose and risk values. The values computed are found to be negligible compared with values from natural background and exposures from nuclear weapon testing.