bluuebunny/crossref_metadata_2025_split huggingface dataset:
Assessment of the look-up table using the tubular and bundle CHF data and modification of the bundle correction factor
Authors: Chandraker D. K. , Vijayan P. K. , Saha D. , Sinha R. K. 
Published: December 2013

Abstract
The Critical Heat Flux (CHF) is an important parameter, which limits the thermal hydraulic performance of the nuclear fuel bundle. The tools available for the prediction of the CHF are empirical in nature and are valid for their experimental range only. However, the recently developed Look-Up Table (LUT) approach has emerged to be a promising tool for predicting CHF in a tubular geometry over a wide range of parameters, which can be extended to the rod bundle geometry considering correction factors for the rod bundle effects. The error statistics of the present assessment confirms the values provided with the LUT for the HBM approach for the tubular application. However, the error statistics by DSM (not provided with the LUT development) is found to be quite different from that of the HBM. It is found that CHF in the rod bundle can also be predicted with a good accuracy using the Heat Balance Method (HBM). The proposed correction factor is found to improve the prediction accuracy of the LUT for the rod bundle application. This paper deals with the assessment of the CHF prediction by LUT for the tubular and bundle geometry and evaluation of the correction factor for rod bundle at the normal operating pressure of BWR (70 bar) using the experimental data base.
A finite element model for static strength analysis of CANDU fuel bundle
Authors: Horhoianu G. , Ionescu D. V. 
Published: December 2013

Abstract
A static strength analysis finite-element model has been developed using the ANSYS computer code in order to simulate the axial compression in CANDU type fuel bundle subject to hydraulic drag loads, deflection of fuel elements and stresses and displacements in the end plates. The validation of the finite-element model has been done by comparison with the out-reactor strength test results. Comparison of model predictions with the experimental results showed very good agreement. The comparative assessment reveals that SEU43 and SEU43L fuel bundles are able to withstand high flow rate without showing a significant geometric instability.
Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects
Authors: Zabunoðlu O. H. , Yavuz U. 
Published: December 2013

Abstract
Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33000, 45000, and 55000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards.
Radiological and thermal characteristics of CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks for spent nuclear fuel storage at Ignalina Nuclear Power Plant
Authors: Poskas P. , Smaizys A. , Simonis V. 
Published: December 2013

Abstract
The main characteristics, such as temperatures of the fuel rod cladding and cask surface, dose rates at the surface and at the some distance for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks loaded with spent nuclear fuel are presented. These casks are used for an interim dry storage of spent nuclear fuel at Ignalina Nuclear Power Plant. Numerical modeling (calculation of the equivalent dose rates, activities of nuclides, etc.) and experimental measurements of the equivalent dose and gamma spectrum on the cask surface at the dry storage facility were performed for assessment of radiation characteristics. Temperatures were evaluated using only numerical modeling. Rather good agreement between experimentally determined and calculated dose rates for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks was obtained. Also it was revealed that maximum fuel rod cladding temperature is higher for CONSTOR RBMK-1500 cask, but never exceeds the maximum allowable value. The cask surface temperatures are similar for both cask types.
Milne problem for isotropic and linearly anisotropic scattering for specular and diffuse reflecting boundary conditions solved with the HN method
Authors: Akgün Y. , Erdogan F. , Tezcan C. 
Published: December 2013

Abstract
In this work, the HN approximation is used to solve the Milne problem for both isotropic and linearly anisotropic scattering. For different specular and diffuse reflection coefficients the extrapolated endpoint is calculated. The numerical results obtained are compared with results already existing in the literature. It is shown that they are in good agreement.
Modelling of the cumulative behaviour of Caesium and Strontium activities in nuclear fuel
Authors: Adlys G. , Adliene D. 
Published: December 2013

Abstract
The nuclear fuel of the operating RBMK-1500 type reactor at Ignalina Nuclear Power Plant is a potential source of accumulated activity which can enter the surrounding environment. Due to the fact, that the high radiotoxic nuclear fuel fission products 90Sr and 137Cs are most important radioactive pollutants, this study is concentrated on the quantitative and qualitative investigation of the cumulative behaviour of all strontium and caesium isotopes during burn-up and cooling time of the nuclear fuel. An investigation is performed on the basis of modelling results, which are important for the risk assessment of environmental contamination and for the interpretation of the results of direct field measurements.
Minimizing radiation exposure for the reactor staff during the dismantling of a TRIGA research reactor
Authors: Lesar M. 
Published: December 2013

Abstract
An experimental method for the determination of the long-term activation of concrete shielding has been performed for two TRIGA research reactors in Ljubljana, Slovenia and in Vienna, Austria. The activity of concrete was studied on concrete samples taken directly from the external shielding of the reactor body with consecutive activation under specified parameters. The present work intends to help minimize radiation exposure for the reactor staff engaged in dismantling the concrete shield of a typical TRIGA Mark II research reactor.
H

N
solutions of the time dependent linear neutron transport equation for a slab and a sphere
Authors: Türeci G. , Güleçyüz M. C. , Tezcan C. 
Published: December 2013

Abstract
The HN method is used to calculate the time eigenvalues in one speed neutron transport theory for a slab and a sphere with given boundary conditions. The angular distribution of the method of elementary solutions and the orthogonality relations of the singular eigenfunctions are used. Calculations are performed considering both the angular and the incoming angular distributions at the boundaries of the given medium. Numerical values are compared with the data available in literature.
National and international standards and recommendations on fire protection and fire safety assessment
Authors: Berg H. P. 
Published: December 2013

Abstract
Experience feedback from events in nuclear facilities worldwide has shown that fire can represent a safety significant hazard. Thus, the primary objectives of fire protection programmes are to minimize both the probability of occurrence and the consequences of a fire. The regulator body expects that the licensees justify their arrangements for identifying how fires can occur and spread, assess the vulnerability of plant equipment and structures, determine how the safe operation of a plant is affected, and introduce measures to prevent a fire hazard from developing and propagating as well as to mitigate its effects in case the fire cannot be prevented. For that purpose usually a comprehensive regulatory framework for fire protection has been elaborated, based on national industrial regulations, nuclear specific regulations as well as international recommendations or requirements. Examples of such national and international standards and recommendations on fire protection and fire safety assessment as well as ongoing activities in this field are described.
Fire tests to validate fire simulation codes: experimental results
Authors: Hosser D. , Riese O. , Dobbernack R. 
Published: December 2013

Abstract
An “International Collaborative Project to Evaluate Fire Models for Nuclear Power Plant Applications” (ICFMP) has been started in 1999 to improve the reliability and the applicability of different fire models with respect to applications for nuclear plants. Within the frame of this project five benchmark exercises (BE) have been conducted from different organizations to represent various fire scenarios. Two of these five benchmark exercises (BE No. 4 and No. 5) have been performed at iBMB in Germany during 2002 and 2004. In BE No. 4 a large pool fire inside a compartment (with a volume of ≈ 76 m3) with respect to under-ventilated situations has been examined in a series of tests. In BE No. 5 four full scale cable tray fire tests have been performed in the same compartment, but with a different construction of the ceiling. The cable tray fire test series with vertically orientated cables considered pre-heating effects and includes time dependent electrical functional failure measurements. In this paper, an overview of the specification and of the experimental results of Benchmark Exercises No. 4 and No. 5 is given.
Fire tests to validate fire simulation codes: numerical results
Authors: Klein-Heßling W. , Riese O. 
Published: December 2013

Abstract
Fire simulations have gained more and more significance, particularly in the context of deterministic as well as probabilistic fire safety analyses for operating nuclear power plants (NPP). An “International Collaborative Project to Evaluate Fire Models for Nuclear Power Plant Applications (ICFMP)” has been started in 1999, to make use of the experience and knowledge of a variety of expert institutions in this field worldwide to assess and improve, if necessary, the state-of-the-art with respect to modelling fires for application to nuclear installations/plants. In the frame of ICFMP two fuel fire experiments have been selected for the Benchmark Exercise No. 4 and four cable tray fire tests for the Benchmark Exercise No. 5, all of these performed in the fire compartment OSKAR of the iBMB (Institut für Baustoffe, Massivbau und Brandschutz) in Braunschweig. In this paper an overview of the calculated results will be given. The simulations of the two large fuel pool fire experiments still indicate difficulties to simulate under-ventilated conditions. The results provide a first indication on the uncertainties of such simulations, although it has to be considered, that the fire load was quite large compared to the compartment volume. The ignition of cables trays as well as the flame propagation on these trays could still not be simulated in an adequate way by the fire codes applied. Different boundary conditions such as pre-heating or the use of more complex cable insulation materials make the situation even more complex.
Probabilistic safety analysis for fire events for the NPP Isar 2
Authors: Schmaltz H. , Hristodulidis A. 
Published: December 2013

Abstract
The “Probabilistic Safety Analysis for Fire Events” (Fire-PSA KKI2) for the NPP Isar 2 was performed in addition to the PSA for full power operation and considers all possible events which can be initiated due to a fire. The aim of the plant specific Fire-PSA was to perform a quantitative assessment of fire events during full power operation, which is state of the art. Based on simplistic assumptions referring to the fire induced failures, the influence of system- and component-failures on the frequency of the core damage states was analysed. The Fire-PSA considers events, which will result due to fire-induced failures of equipment on the one hand in a SCRAM and on the other hand in events, which will not have direct operational effects but because of the fire-induced failure of safety related installations the plant will be shut down as a precautionary measure. These events are considered because they may have a not negligible influence on the frequency of core damage states in case of failures during the plant shut down because of the reduced redundancy of safety related systems. The modelling of the fire-induced influences in the PSA-project also includes, in extension to the existing PSA for power operation, components which were needed for full power operation. The fire-induced failure of these components can result in transient events. To be able to assess the influences of fire events on the plant safety it is necessary to have detailed information about the arrangement of the components which were modelled in the PSA and their cable routing. This information has to be integrated into the PSA model. With help of this comprehensive and detailed modelling of fire specific influences on operational and safety systems a PSA model was developed, which allows to assess and to point out fire-induced influences on component level. This proceeding provides the opportunity to carry out measures to optimize the plant design and thus to enhance the plant safety.
The revised program for measurements in intense operation mode according to AVV-IMIS
Authors: Bieringer J. , Bühling A. , Haase G. , Heinrich T. , Müller-Neumann M. , Steinkopff T. , Wiezorek C. , Wirth E. 
Published: December 2013

Abstract
The monitoring program for measurements in intense operation mode has been revised recently by a working group at the Federal Ministry for Environment, Nature Conservation and Reactor Safety (BMU). The major issues of the revision are reported in this contribution. Measurements in intense operation must be appropriate for fast assessment of the radiological situation, for estimating the dose to the population and for decisions on countermeasures to minimize the dose. In order to meet these requirements the structure of the measurement program in intense operation mode was divided into three phases when different exposition paths are relevant: before and during dispersion of radioactive material, immediately after dispersion of radioactive material has ended and a late phase when contamination values have decreased in different environmental media. For each of these phases a special measurement program was defined that is tailored to achieve the above mentioned objectives. Minimum detectable activity concentrations were introduced similar to the measurement program in routine operation mode. They follow the intervention levels in the catalogue of countermeasures and maximum permitted values given by the European Union (EU) for food and animal feed. The minimum detectable activity concentrations were defined such that the detection of 1/10 of the intervention levels for countermeasures is ensured.
Early emergency response by means of dispersion forecasting – emergency management of the Deutscher Wetterdienst in the context of national and international agreements
Authors: Steinkopff T. , Fay B. , Glaab H. , Jacobsen I. , Klein A. , Mirsch M. 
Published: December 2013

Abstract
Based on results of the numerical weather forecast models; the DWD (Deutscher Wetterdienst, German Meteorological Service) is immediately able to calculate dispersion forecasts in the case of an emergency with an emission of radioactive inventory. This information is essential for the management of further countermeasures. There are legal agreements between the federal states of Germany and the Federal Government regulating the mutual aid. For the emergency preparedness system RODOS (Real Time Online Decision Support) the results of the numerical weather forecasts are provided and complemented by medium-range dispersion forecasts. In the framework of the German computer network IMIS (Integrated Measurement and Information System for the Surveillance of Radioactivity in the Environment) the results of trajectory calculations are presented graphically within 10 minutes, those from the LPDM (Lagrangian Particle Dispersion Model) forecasts within 30 minutes. In case of a nuclear emergency, global distribution of information via the WMO (World Meteorological Organization) communication system GTS is part of international agreements between IAEA (International Atomic Energy Agency) and WMO, thus giving the DWD a key role. The commitments of the DWD in managing nuclear emergencies in the framework of national and international agreements are presented, as well as internal emergency arrangements.
Implementation of decision support systems in Austria
Authors: Fritsch G. , Haider W. , Hofer P. , Hohenberg J.-K. , Karg V. 
Published: December 2013

Abstract
According to the Austrian Radiation Protection Act, decision support systems have to be implemented and operated in Austria. Recommendations on early counter measures in case of a radiological emergency are based on the results of these decision support systems and on measurements of the environmental monitoring. The following article gives an overview on the different decision support systems operated in Austria and first experiences gained through their application in international exercises and projects.
Longitudinal dispersion of radioactive substances in Federal waterways
Authors: Krause W. J. , Speer W. , Lüllwitz T. , Cremer M. , Tolksdorf W. 
Published: December 2013

Abstract
In the context of radioactivity monitoring in German Federal Waterways (BWStr) by the Federal Institute of Hydrology (BfG) according to the Precautionary Radiation Protection Act (StrVG), the prediction of the dispersion of radioactive substances in water is one of the key tasks. The aim is the forecasting of the longitudinal dispersion of concentrations of soluble hazardous substances in flowing water. These predictions are based on the so-called dispersion tests with tritium as a tracer that the BfG has performed since 1980. Characteristic parameters like discharge-dependent flow velocities, dispersion and elimination constants related to emission sources or selected river sections are determined. They will serve as basis for a mathematical model to forecast discharge-dependent flow velocities, expected impact times, concentration maxima, and the duration of critical increases in concentrations. In the following, the results obtained till now from three investigation campaigns on the River Weser and its source rivers Werra and Fulda are described.
The extension of the 2-D finite element/boundary element hybrid method to general multigroup neutron diffusion theory
Authors: Ozgener B. , Cavdar S. , Ozgener H. A. 
Published: December 2013

Abstract
The finite element-boundary element hybrid method developed previously for reflected systems and restricted to one or two group neutron diffusion theory has been extended to the general multigroup neutron diffusion theory by using the boundary integral equation of multigroup neutron diffusion theory. A linear or bilinear 2-D FEM formulation in the core combined with a 2-D linear BEM formulation in the reflector constitute the basic discretization procedure. Use of the boundary integral equation of multigroup diffusion theory transforms all group-to-group scattering domain integrals into surface integrals in the reflector. Hence the need for a reflector domain mesh is completely eliminated. Via comparisons with pure FEM and BEM solutions of the reflected systems within the context of three and four group diffusion theories, the present formulation is validated and assessed.
Extension of the HN
method for solving radiation transfer problems in plane geometry for general anisotropic scattering
Authors: Tezcan C. 
Published: December 2013

Abstract
The HN method is used to solve radiation transfer problems for plane parallel media for anisotropic scattering. This approach is based on the use of the angular distributions in terms of singular eigenfunctions of the method of elementary solutions. Two analytical expressions are derived for the coefficients of the power series expansions in exit distributions.
Studies on the transmission and processing of pulse-shaped signals from nuclear radiation detectors using methods of systems theory
Authors: Spillekothen H.-G. 
Published: December 2013

Abstract
Using methods of the systems theory of electronic communications and theoretical electrical science, this study describes the transmission of pulse-shaped signals from nuclear radiation detectors from the detector over “electrically long lines” (cables) to the output of the first pulse amplifier. The example of pulses from BF3-proportional counters shows, using the Fourier transformation, that pulses from radiation detectors contain a frequency spectrum ranging well above 108 Hz. If these pulses are transmitted to the first amplifier over a line length of several meters, the laws of the theory of transmission lines must be taken into account to avoid false signals caused by reflections. In the example, line equations are applied and the influence of the line and the terminating impedance is demonstrated. The influence of the frequency response v(ω) and the phase response δ(ω) of the amplifier is also considered in the sample calculation. The methods presented make it possible to analyze and optimize the transmission and amplification of signals from radiation detectors. Close agreement emerges between empirically observed and calculated pulse shapes.
The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
Authors: Öztürk H. 
Published: December 2013

Abstract
The Chebyshev polynomial approximation is used to solve the reflected critical slab problem for one-speed neutrons in a slab with strongly anisotropic scattering. The scattering kernel which is a combination of backward-forward-isotropic scattering and linearly anisotropic scattering is chosen in a uniform finite slab. The critical slab thicknesses are given for different degrees of reflection, backward-forward and linear anisotropy. Calculated numerical results for the critical thickness are in good agreement with the results available in literature.
Dynamic structure response due to reactor cooling piping system failure
Authors: Ahmed K. , El-messiry A. M. 
Published: December 2013

Abstract
The safety evaluation of the primary core cooling system requires an accurate hydraulic analysis of the system during normal and abnormal conditions. A model of the piping components of a common research reactor system is obtained and illustrated using the FE/PIPE Finite Element Package. Different accidents leading to a loss of the system function are investigated, such as pipe rupture and valve failure. The hydraulic-structure interaction and dynamic parameters responded to the accident are obtained during the transient time. The scenario time domain is selected to be very short (2.0 s) to get more conservative calculations and more safety margins in the final design. The results reflect that the natural system frequency should be in the margin greater than 34 Hz otherwise resonance with the generated forces during the accident will happen, that could lead to a system failure.