bluuebunny/crossref_metadata_2025 huggingface dataset:
Validation of the CFD code FLUENT by post test calculation of a ROCOM experiment with density driven coolant mixing
Authors: Schaffrath A. , Fischer K.-C. , Hahm T. , Wussow S.
Published: December 2013

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
Authors: Weil L.
Published: December 2013

The German regulatory PSA guide describes the fundamental requirements with respect to probabilistic safety assessments as a central element of comprehensive safety reviews. The technical details regarding the performance of PSA are set out in two technical documents (PSA Methods and PSA Data), which have been developed by a working group of PSA experts (FAK PSA). Based on the experiences gained from the first series of periodic safety reviews in Germany, on international experience and on the fact that comprehensive safety reviews are now mandatory since April 2002, all guides hat to be revised. The first guide updated was the PSA guide together with the corresponding technical documents, issued October and November 2005. This article describes the key points of the revised requirements on PSA methods and data as well as their associated potential to improve emergency planning on nuclear sites.
Determination of the RBMK-1500 reactor passport characteristics
Authors: Pabarcius R. , Tonkunas A. , Listopadskis N.
Published: December 2013

During the whole life of reactor equipment it has to be shown that safety relevant limitations of general reactivity characteristics are not exceeded. Usually, safety margins of these characteristics are listed in a technical certificate (reactor passport). This paper presents the calculation results of the Ignalina nuclear power plant unit 2 reactor passport characteristics obtained by the best estimate 3D computer code QUABOX/CUBBOX, adjusted for the RBMK-1500 reactor core. The characteristics determined are supported by an uncertainty and sensitivity analysis which was performed using the software package SUSA. The uncertainty of the reactivity coefficients and associated effects were quantified in terms of statistical tolerance limits. The sensitivity analysis provided a ranking of the uncertain input parameters with respect to their contribution to the uncertainty of the reactivity characteristics.
Application of the spectral Green's function method for numerically solving discrete ordinates problems in cylindrical geometry
Authors: Öztürk H. , Anlı F. , Güngör S.
Published: December 2013

The spectral Green's function (SGF) method is used to solve numerically the transport of neutrons in an infinite homogeneous cylindrical domain. In the method, the scattering of neutrons is assumed to be isotropic and the transport equation in cylindrical geometry is reduced to plane-like transport equation by using our reasonable approach. Then, the plane-like transport equation is solved by using the SGF method and the results are compared with those already obtained by the Diamond Difference (DD) scheme to check the accuracy of the results. The agreement of the results is good and it can be concluded that the SGF method is very effective for the numerical solution of the transport equation.
Phenomenological model of TENORM formation in industrial processes
Authors: Gellermann R.
Published: December 2013

This paper discusses a simple phenomenological model of TENORM formation. The model is based on a mass and activity balance in an industrial process. It characterises the radionuclide enrichment of a residue in relation to the input materials according to three parameters (mass transfer factor MTF, activity transfer factor ATF and enrichment factor EF). The model shows that high enrichment is only possible if mass transfer into the residue is low. By analysing data from several real industrial processes, the co-precipitation of dissolved radium with barium salts (EF about 104), the volatilisation of lead and polonium in thermal processes (EF about 300) and the combustion of fuels (EF 10–100) are identified as very effective enrichment processes. However, in real industrial processes the specific activity of residues may be influenced by admixtures of masses after the decisive step of radionuclide redistribution. Therefore, focus should be placed on the identification and characterisation of elementary processes of TENORM formation. Several such processes are discussed in this paper.
Development of an object-specific remediation solution for the waste rock dumps at the formerly abandoned Wismut sites
Authors: Regner J. , Schulz H.
Published: December 2013

The investigation of the radiation situation is an essential part of preparing the remediation of the formerly abandoned Wismut sites. If the exposure is relevant, then the exposure pathway “inhalation of radon/radon daughters” plays the dominant role. Therefore, the development of a straightforward approach suitable for the evaluation of the involved radon situation is one of the most important tasks. It has to include the whole chain beginning with the characterization of the relevant exposure situation and ending up with the identification of remedial options. For that purpose a new technique of visualizing the radon flux from the waste rock dumps has been developed. The specific conditions at formerly abandoned Wismut sites will be demonstrated and the remediation solutions envisaged to reduce the radon flux will be explained.
Radon in Austria: metrology and practice
Authors: Maringer F. J.
Published: December 2013

The implementation of radon mitigation and precaution standards needs continuously scientific attendance and research networking on international level. Otherwise the radon issue could degrade easily to a simplified techno-economical exercise without sustainable results in public health. In this paper the radon investigations in Austria which have been carried out in the last 20 years and the applied methods and derived standards for mitigation and precaution at home and workplaces are discussed. Future strategies, scientific and social necessities to solve the radon problem are outlined comprehensively. The agreed and proposed future research objectives and cooperations in Austria that seem absolutely necessary are discussed in consideration of the medium-term perspective of the European radiation protection.
Workplaces with natural radionuclides – investigation results from Bavaria
Authors: Reifenhäuser C. , Körner S. , Mallick R.
Published: December 2013

Since the amendment of the German Radiation Protection Ordinance in the year 2001 special workplaces with increased exposure due to natural terrestrial radiation have been subject to legal regulations. After the Radiation Protection Ordinance came into force the Bavarian Environment Agency informed all water supply facilities, breweries, dairies, mineral water manufactures, operators of underground workplaces and radon spas that the annual radon exposure of their staff has to be estimated. In Bavaria approximately 3330 companies with workplaces with exposure to radon were investigated. Herein the water supply facilities represented the largest group with 2550 companies. In about 96 % of all investigated companies the radon exposure of the staff is below the action level of 6 mSv per year. In 75 water supply facilities and two mines the radon exposure of the staff exceeded the action level (6 mSv/a) or the annual limit (20 mSv/a). Workplaces with increased exposure due to uranium and thorium are of secondary importance in Bavaria. Within the scope of a research project 13 companies were found, that still use thoriated welding electrodes. Only a few facilities use thoriated mantles, thorium alloys or natural thorium and uranium. The effective dose of all examined workers was below the action level of 6 mSv per year.
Spherical albedo problem for isotropic scattering
Authors: Türeci R.G.
Published: December 2013

The pseudo-slab problem by using the HN method, proper neutron flux definitions and physical boundary conditions is used to study spherical albedo. The selected numerical values can be compared with data available in the literature. The sphere radius can be calculated for the secondary neutron number and for the spherical albedo.
Study on the thermal-hydraulics characteristics of a boiling two-phase natural circulation loop with nanofluids
Authors: Nayak A. K. , Gartia M. R.
Published: December 2013

In this paper, we present the steady state and stability behaviour of a boiling two-phase natural circulation loop with a small concentration of Al2O3 nanofluids. For this, experiments were conducted in a natural circulation loop at different operating powers, pressures, and water levels in the steam drum. The test results revealed that while the steady state natural circulation behaviour is similar, however, the buoyancy induced flow rates are found to be relatively higher with nanofluids than with water alone. In addition, with nanofluids, the amplitudes of the boiling induced Type I instabilities are found to be significantly reduced, However, the Type II instabilities are found to appear at relatively lower power with addition of nanofluids as compared to that with water alone. The most interesting result is that with addition of nanofluids, the amplitudes of Type II instabilities are found to get suppressed with increase in power similar to that of low quality Type I instability.
The effect of combination of different materials on neutron absorption in a nuclear research reactor spent fuel pool
Authors: Logar M.
Published: December 2013

So far, only individual absorbing materials have been studied for neutron absorption in the pool-type reactor spent fuel storages. In this paper, the effect of combining various materials, which are already known as good neutron absorbers by themselves, on lowering the multiplication factor of spent fuel pool, is described. The results of Monte Carlo computer code MCNP calculations for the spent fuel pool of the TRIGA research reactor in Ljubljana, Slovenia, are specified. Due to property completion of the chosen absorber materials, some of their combinations partially decrease the multiplication factor and thus contribute to the enhancement of the safety conditions of the pool.
Re-evaluation of the criticality experiments of the “Otto Hahn Nuclear Ship” reactor
Authors: Lengar I. , Snoj L. , Rogan P. , Ravnik M.
Published: December 2013

Several series of experiments with a “FDR” reactor (advanced pressurized light water reactor) were performed in 1972 in the Geesthacht critical facility ANEX. The experiments were performed to test the core prior to its usage for the propulsion of the first German nuclear merchant ship “Otto-Hahn”. In the present paper a calculational re-evaluation of the experiments is described with the use of the up-to date computer codes (Monte-Carlo code MCNP5) and nuclear data (ENDF/B-VI release 6). It is focused on the determination of uncertainties in the benchmark model of the experimental set-up, originating mainly from the limited set of information still available about the experiments. Effects of the identified uncertainties on the multiplication factor were studied. The sensitivity studies include parametric variation of material composition and geometry. The combined total uncertainty being found 0.0050 in keff, the experiments are qualified as criticality safety benchmark experiments.
Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti
Authors: Horhoianu G. , Dobrea D. , Parvan M. , Stefan V.
Published: December 2013

Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g–1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti.
Effect of using various grades of plutonium in the protective liquid wall of an IFE type fusion reactor
Authors: Übeyli M.
Published: December 2013

One of the main issues in a fusion reactor is the need for the frequent replacement of first wall structures during its lifetime. In order to eliminate this problem completely, use of a protective liquid wall between the fusion plasma and solid first wall has been proposed by various researchers. One of the major fusion energy reactor designs is called HYLIFE-II, an inertial fusion energy (IFE) type reactor that also utilizes such a flowing liquid wall to extend the lifetime of the first wall to its lifetime. This study presents the neutronic analysis of the HYLIFE-II reactor using Li2BeF4 (Flibe) with various grades of plutonium as a flowing liquid wall. For this purpose, four different coolants containing Flibe with 0.5 % or 1 % of weapon grade (WG) PuF4 and Flibe with 0.5 % or 1 % reactor grade (RG) PuF4 were considered. Main neutronic parameters: tritium breeding ratio (TBR), energy multiplication factor (M) and radiation damage at the first wall were computed with respect to the liquid wall thickness and coolant type. The highest TBR and M values were reached when the Flibe with 1 % WG PuF4 was used in the liquid wall.