bluuebunny/crossref_metadata_2025 huggingface dataset:
Uncertainty and sensitivity analysis of void and power reactivity coefficients in an RBMK-1500 reactor core
Authors: Pabarcius R. , Tonkunas A. , Bubelis E. , Clemente M.
Published: December 2013

Uncertainty and sensitivity analyses were performed for the calculation of power and void reactivity coefficients of an RBMK-1500 reactor core state. Both parameters are important for the reactivity feedback behaviour and are of special safety relevance. The reactivity coefficients were calculated using the 3D-reactor core model QUABOX/CUBBOX-HYCA and the uncertainty and sensitivity analyses were performed by the GRS methodology applying the SUSA software package (Software System for Uncertainty and Sensitivity Analysis). A total number of 20 uncertain input parameters and their probability distributions were identified as possibly relevant for the size of the reactivity effects and taken into account in the SUSA analysis. The method helps to identify the main contributors to the uncertainty of calculated results, as well as to determine two-sided (95 %, 95 %) statistical tolerance intervals of the investigated reactivity coefficients. The approach to quantify uncertain parameters for RBMK-1500 core calculations using the SUSA code package and QUABOX/CUBBOX-HYCA was successfully performed. The results provide guidance where model improvements should be performed to increase the accuracy of the calculations efficiently.
Reactivity worth of the control rods at the ET-RR-2 reactor core
Authors: Saleh H. G. , Ashoub N.
Published: December 2013

Safety analysis of research reactors include a determination of control rod worth and kinetic parameters upon which the dynamic performance of the reactor depends. The core control of the ET-RR-2 reactor is accomplished through six Ag-In-Cd alloy-absorbing plates, which can be inserted into the core with a high velocity. In the present work, the fully and partially inserted control rod configuration in 3D reactor geometry and the effect of reactivity worth of the control rods on the reactor core is studied. The reactivity worth were calculated for each control rod and also for combinations of more than one rod. The total worth of all control rods in the reactor core is about 16.95% k/k from the core reactivity. Other neutronic parameters, i. e. the 3D neutron flux distribution and the change of the power peaking factor as a results of inserted and withdrawing the control rods were calculated.
Criticality analyses of regions containing uranium in the earth history
Authors: Ravnik M. , Jeraj R.
Published: December 2013

Investigations of necessary conditions for a self-sustained chain reaction in the Earth inner regions hypothetically containing uranium is presented for the time interval from Earth formation to present time. It is determined that criticality was theoretically possible up to 2.5 Ga before present if uranium concentrated in pure form. In the early geological history (4 Ga before present) the self-sustained criticality could occur even if uranium was diluted up to 1: 20 by the average core material or 1: 60 by the average mantle material. If other metallic materials of similar density as uranium (e. g., Au, W) or similar atomic weight (e. g., Th) concentrated from the primordial mixture in equal proportion as uranium, criticality was not possible for any period in Earth history provided that the basic material contained no light nuclides (H, C). Criticality in the Earth inner regions could have established only if uranium concentrated from the basic material more effectively than elements of similar density or atomic number.
Brennstabauslegung und Brennstabmodellierung – Teil 2
Authors: von Collani Elart , Baur Karl
Published: December 2013

Im Teil 1 dieser Untersuchungen wurde in das Problem der Brennstabauslegung eingeführt sowie die Grundlagen für das Verständnis des zweiten Teils gelegt. Dieser zweite Teil beschäftigt sich mit der mehr technischen Seite der mathematischen Modellierung des Verhaltens der Brennstäbe im Einsatz. Zusätzlich wird auf die in Deutschland aktuelle Diskussion über das sogenannte Ein-Brennstab-Quantil eingegangen.
The chemical separation of Eka-Hg from CERN W targets in view of recent relativistic calculations
Authors: Kolb D. , Marinov A. , Newton G. W. A. , Brandt R.
Published: December 2013

In 1971 evidence for the production of element 112 via secondary reactions in CERN W targets was obtained. The evidence was mainly based on the observation of fission fragments in Hg sources separated from the W targets, on the measured masses of the fissioning nuclei and on the assumption that element 112 (Eka-Hg) actually behaves like Hg in the chemical separation process. This assumption is analyzed in view of recent relativistic calculations of the electronic structure of element 112. It is shown that in the superheavy element region only the chemistry of element 112 is similar to that of Hg.
Is mapping a part of common cause failure quantification?
Authors: Vaurio J. K.
Published: December 2013

Important issues in any procedure used for estimating basic event probabilities of common cause failures (CCF) for probabilistic safety assessments (PSA) are: which plants and systems to use, how to combine them, and how to transform data from systems with different numbers of similar components to obtain CCF-rates for a specific group of components. These issues are addressed with focus on the last part called “mapping”. Certain parametric models are considered for transforming CCF event experience from data-source plants to the target plant, the plant of interest. Two sets of rules are reviewed and compared for transforming rates and assessment uncertainties from larger to smaller systems i.e. mapping down. Epistemic uncertainties are taken into account in the estimation. Mapping down equations are presented also for the alpha-factors and for the variances of CCF-rates. Consistent rules are developed for mapping up CCF-rates, uncertainties and alpha factors from smaller to larger systems. These rules are not limited to a binomial CCF model. Consistency requirements are severe and dictate certain limits to possible parametric values. Empirical alpha factors are used to quantify robust mapping ratios of complete CCF-rates. Mapping is critically analyzed and practical recommendations are made.
Experiments and comparing calculations on thermohydraulic pressure surges in pipes
Authors: Neuhaus T. , Dudlik A.
Published: December 2013

Modelling of pressure surges in power plant pipe systems means to carefully judge and classify the relevant variables. A decision must be made, whether the numerous effects that can occur during a transient event, should be modelled. One main point is to determine if fluid structure interaction (FSI) is important for the special situation/scenario. Also unsteady friction and degassing of dissolved gases can occur. But with each model more sources of errors are introduced into the mathematical system, such as uncertainty about bearing clearance, determination of fixing points/stiffness/spring constants, determination of friction coefficients and damping parameters, uncertainty about the amount of non-condensables in the liquid, dynamics of de-aeration processes and unsteady friction coefficients. Another complicated problem is the definition of boundaries, like specifications of pumps, valves, fittings, etc. It is obvious that a combination of all uncertainties may lead to quite uncertain predictions and must therefore be very carefully taken into account. Within this contribution, valve-closing experiments of different initial temperature and system pressure at Fraunhofer test rig PPP are presented. Two scenarios are modelled using different assumptions and model depths and the simulation results are compared with experimental data.
Irradiation damages in tantalum spallation target
Authors: Villagrasa-Roussel C. , Broeders C. H. M. , Konobeyev A. Yu.
Published: December 2013

The irradiation by a high intensity proton beam of the spallation target in Accelerator Driven Systems (ADS) gives rise to structural damages that must be considered in the design and handling of such materials. In this work, some of these damages are evaluated in the case of the irradiation of a tantalum target. Tantalum was recently proposed as solid target in the case of low power systems. Unfortunately, not enough experimental data of spallation reactions in this material are available. Therefore, our calculations were made using MCNPX and the stand-alone version of the spallation codes INCL4+ABLA. Results of this work are the concentrations of spallation residues produced in the target after one year of irradiation and calculations of the activity derived. Also displacements per atom in the target material have been evaluated. Different components (elastic scattering and non-elastic reactions) of the displacement cross-sections were calculated with two different methods.
An interpretation procedure for the purpose of incorporation monitoring during decommissioning of nuclear reactors
Authors: Schupfner R. , Haas G. , Wolfbeis O. , Dettmann K.
Published: December 2013

The incorporation monitoring during decommissioning of nuclear reactors is faced with a complex situation arising from a mixture of radionuclides. The development of an interpretation procedure to enable the application of an optimised incorporation monitoring on conditions of decommissioning of nuclear reactors has been carried out combining results from radionuclide analyses of contaminated or activated materials and the proven methods of incorporation monitoring in the case of the intake of single radionuclides. In this study the radiological requirements, the mathematical elements of the interpretation procedure are presented as well as the application to an activated concrete of biological shielding.
Thermal hydraulics of the fixed bed nuclear reactor concept
Authors: Sefidvash F.
Published: December 2013

The thermo-hydraulics of the fixed bed nuclear reactor (FBNR), possessing spherical HTGR type fuel, is described. Its main characteristic is the variable fixed bed core in suspense and its “cold” fuel in case of a LOCA (< 400°C). A 40 MWel reactor has a pressure loss across the fixed bed of 9.5 bar and needs a pump power of 3.4 MWel – more than in a classical PWR but being the price for its excellent inherent safety characteristics.
The critical slab problem for linearly anisotropic scattering and reflecting boundary conditions with the HN method
Authors: Güleçyüz M. Ç. , Türeci R. G. , Tezcan C.
Published: December 2013

The HN method is used to calculate the critical slab thickness for a parallel media surrounded bay a reflector for linearly anisotropic scattering. It is shown that the method leads to concise equations and accurate numerical results even in the lowest order of approximation.
The multipurpose thermalhydraulic test facility TOPFLOW: an overview on experimental capabilities, instrumentation and results
Authors: Prasser H.-M. , Beyer M. , Carl H. , Manera A. , Pietruske H. , Schütz P. , Weiß F.-P.
Published: December 2013

A new multipurpose thermalhydraulic test facility TOPFLOW (TwO Phase FLOW) was built and put into operation at Forschungszentrum Rossendorf in 2002 and 2003. Since then, it has been mainly used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of Computational Fluid Dynamic (CFD) codes in the frame of the German CFD initiative. The advantage of TOPFLOW consists in the combination of a large scale of the test channels with a wide operational range both of the flow velocities as well as of the system pressures and temperatures plus finally the availability of a special instrumentation that is capable in high spatial and temporal resolving two phase flow phenomena, for example the wire-mesh sensors.
Effect of packing fraction variations on the multiplication factor in pebble-bed nuclear reactors
Authors: Snoj L. , Ravnik M.
Published: December 2013

The effect of fuel element packing density on the effective multiplication factor, keff, in a typical pebble-bed reactor (PBR) is studied using MCNP and WIMS computer codes. It is observed that at normal operating conditions keff increases with packing density. Increase in packing density of 1% results in approximately 0.0015 increase in keff. A negative temperature effect is observed at all feasible packing densities. The effect of secondary coolant (water) ingress in the core at accidental conditions is studied at various packing densities. The effect of effective water density on keff depends on packing fraction, fuel enrichment and fuel-to-graphite ratio. It is strongly positive in highly enriched fuel at low water density regardless of packing fraction. For fuel enrichment 8% and normal packing density, a maximum positive effect of approximately +0.10 on keff is observed at a water density of ∼0.1 g/cm3. The positive effect is reduced at low enrichments. Above a water density of 0.5 g/cm3, the effect on keff becomes negative regardless of enrichment and packing density.
Modeling of the TOSQAN test facility with the lumped parameter code COCOSYS
Authors: Babilas E. , Urbonavicius E. , Rimkevicius S.
Published: December 2013

Hydrogen distribution in the containment during severe accidents is an important safety issue of nuclear power plants operation. The detailed knowledge of the containment thermal-hydraulics is necessary for reliable predictions of pressure, temperature and gas distribution in the containment of the NPP during postulated accidents. For this purposes, the study of the International Standard Problem (ISP 47) has been started in 1999 with the test facilities TOSQAN, MISTRA and ThAI. The paper presents the COCOSYS nodalisation of the TOSQAN test facility. The results are analysed and compared with the experimental values.
Analytical and experimental investigations of shear stress in rod bundles with irregular cells
Authors: Cesna B.
Published: December 2013

The paper presents analytical and experimental investigations of shear stress on the rods of a bundle with irregular arrays. Simplified, however, effective methodologies are developed for thermal hydraulic processes (shear stress on the bundle rods' surfaces) along the multi-rod bundles. A single-vortex transfer gradient model is recommended taking into account the calculation results and a number of referenced experimental works. Analytical calculation methodologies of shear stress are based on ordinary differential equations and implemented in the computer codes RKN and RKN-M. Concurrently with the analytical investigations, experimental works were performed to determine the shear stress on the rods in a bundle with irregular cells. The results of the calculations are compared with experimental data.
Time-dependent albedo problem for quadratic anisotropic scattering
Authors: Türeci R. G. , Türeci D.
Published: December 2013

Case's eigenfunctions and their orthogonality relations can be improved assuming pure-quadratic anisotropic neutron scattering. As a result, the infinite medium Green function can be written with those eigenfunctions terms. Choosing the appropriate boundary conditions and using the method of singular eigenfunctions, the half-space values are obtained. Our numerical results for this problem are reported.
European approach to nuclear and radiological emergency management and rehabilitation strategies (EURANOS)
Authors: Raskob W.
Published: December 2013

The 5-year multi-national project EURANOS, funded by the European Commission and 23 European Member States, started in April 2004. Integrating 17 national emergency management organisations with 33 research institutes, it brings together best practice, knowledge and technology to enhance the preparedness for Europe's response to any radiation emergency and long-term contamination. Key objectives of the project are to collate information on the likely effectiveness and consequences of a wide range of countermeasures, to provide guidance to emergency management organisations and decision makers on the establishment of an appropriate response strategy and to further enhance advanced decision support systems, in particular, RODOS (Real-time On-line Decision Support), through feedback from their operational use. The research activities focused on emergency actions and countermeasures, the enhancement of decision support systems for operational application and the development of strategies for the sustainable rehabilitation of contaminated territories. The demonstration activities exercise the developed methods and tools in the actual operational environment.
Bilateral information and data exchange in case of nuclear emergencies in the German-Dutch border region
Authors: Twenhöfel C. J. W. , Salfeld H. Ch. , Hable K. , Reinen H. A. J. M.
Published: December 2013

Communication and exchange of accident information and monitoring data during nuclear emergencies is of crucial importance for a coherent crisis management, especially in cross border regions. In Germany and the Netherlands the need for a reliable information and data exchange for cross border emergency management was recognized and resulted in a Memorandum of Understanding between our countries. A (sub) working group of the Niederländisch-Deutsche Kommission für grenznahe kerntechnische Einrichtungen (NDKK) was formed by NLWKN, BfS and RIVM. The working group reviewed the technical requirements and operational aspects of a coherent data and information exchange between national and federal emergency centers and the state authorities. The data exchange comprises various messages, documents, measurement data and prognostic results from model calculations from the technical decision support centers of NLWKN, BfS and RIVM. Integrated accident information is available on the emergency management websites ELAN (D) and CalWeb (NL). Via regular data exchange programs and exercises special attention is given to ensure long term sustainability of the technical infrastructure and operational data exchange procedures.
Rapid determination of strontium radionuclides in plants, fodder and foodstuffs
Authors: Tait D. , Haase G. , Hartmann R. , Jelinski M.
Published: December 2013

The fission yield, the transfer factors in the food chain and the dose coefficient are large for the nuclear fission product Sr-90. The surveillance of Sr-90 in the food chain is therefore important in precautionary radiation protection and in assessing the radiation dose to the public especially after a nuclear incident. Prior to analysis, as it is a pure β-emitter, Sr must be separated from the sample by procedures which, for complex organic samples, are lengthy, laborious and dependent on operator skill. Ubiquitous natural radionuclides and short-lived fission products in samples contaminated with fresh fallout may interfere. Here we describe a fast, reproducible and effficient method for extracting Sr from grass, clover, maize, wholemeal rye, baby food, and total diet. The method depends on obtaining an ash free of traces of organic interferences. Sr may be separated from a dilute nitric acid leachate of such ash with a solution of dicyclohexyl-18-crown-6 in chloroform. Interfering radionuclides are removed with a special manganese (IV) oxide (active, precipitated from Merck). Sr is precipitated as carbonate then dispersed in a cocktail for liquid scintillation spectrometry. This allows simultaneous counting of Sr-89 (a short-lived β-emitter in fresh fallout) and Sr-90. The chemical yields of Sr determined with the gamma-emitting Sr-85 tracer are reproducible and greater than 75 % in all cases. The sample ashing requires 18 h and the extraction 4.0 to 4.5 h. Thus, a duplicate analysis may be completed within 2 days of receipt of the sample.
Correction of deposition predictions with data assimilation
Authors: Gering F.
Published: December 2013

Model predictions for rapid assessment and prognosis of possible radiological consequences after an accidental release of radionuclides play an important role in nuclear emergency management. Radiological measurements (e. g., dose rate measurements) can be used to improve such model predictions. This paper describes a method for combining model predictions and measurements (data assimilation) in the deposition model of the European radiological decision support system RODOS. The data assimilation approach is based on the Ensemble Kalman Filter, a Monte Carlo variant of the Kalman filter.
Iodine Prophylaxis following nuclear accidents – a concept how to distribute potassium-iodide tablets out of the central stocks in the event of an accident
Authors: Portius U.
Published: December 2013

With its recommendation “Iodine prophylaxis following nuclear accidents” (1996) and its reports of 1997 and 2001 the German Commission on Radiological Protection (SSK) followed the recommendations of the WHO “Guidelines for iodine prophylaxis following nuclear accidents” of 1989. The intervention levels were lowered (50 mSv for children/adolescents (up to the age of 18 years) and pregnant women, 250 mSv for adults), the iodine prophylaxis was restricted to persons up to the age of 45 years and the recommended dosage of stable iodine was changed. Due to the lowered reference levels the radius of 25 km around a nuclear power plant that had been the planning radius for the distribution of iodine tablets so far was extended to 100 km. Based on these recommendations the German authorities began to set up new strategies for the provision and distribution of potassium-iodide tablets (iodine tablets). Since 2004, within the radius of 25 km the iodine tablets are pre-distributed to households and/or stored at several points in the municipality for persons up to the age of 45 years. For the new planning radius of 25 – 100 km iodine tablets are stored in 8 central stocks in Germany for children/adolescents (up to the age of 18 years) and pregnant women. A working group with representatives from federal and Länder authorities has developed a distribution strategy for the distribution out of these central stocks in the event of an accident. It describes a possibility of organising and implementing the distribution of the iodine tablets within the radius of 25 – 100 km in a nationwide standardised way.
The slab albedo and criticality problem for the quadratic scattering kernel with the HN method
Authors: Türeci R. G. , Güleçyüz M. Ç.
Published: December 2013

The one speed, time independent neutron transport equation for slab geometry with quadratic anisotropic scattering kernel is considered. Albedo, transmission factor and criticality thickness are calculated by the HN method. The numerical results obtained are listed for different selected parameters. It is shown that the method is concise and leads to fast converging numerical results.
Study of the effect of anisotropic scattering on the critical slab problem in neutron transport theory using Chebyshev polynomials
Authors: Öztürk H.
Published: December 2013

The effect of anisotropic scattering on the variation of the critical thickness for one-speed neutrons in a uniform finite slab has been studied using Marshak boundary conditions. Chebyshev polynomials of the second kind are used in the angular part of the neutron flux. The critical slab thicknesses are calculated for various values of collision and anisotropy parameters. Numerical results obtained by the present method for the critical thickness are presented in the tables together with the results obtained by Legendre polynomials approximation and the ones available in the literature for comparison.
Analysis of a steam generator tube rupture under realistic initial and boundary conditions
Authors: Sarkadi P. , Schaffrath A.
Published: December 2013

Within the framework of safety analysis the TÜV NORD SysTec GmbH & Co. KG has analyzed a steam generator tube rupture for a pressurized water reactor under realistic initial and boundary conditions. During the accident an undesirable backing of the coolant pressure by switching-on the safety injection pumps is avoided, which can endanger the coolant retention in the steam generators, because the shut off heads of the safety injection pumps are larger than the response pressure of the safety valves.
Application of the Laplace transform method for the albedo boundary conditions in multigroup neutron diffusion eigenvalue problems in slab geometry
Authors: Petersen C. Z. , Vilhena M. T. , Barros R. C.
Published: December 2013

In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme.
Training and retraining programme for research reactor operating personnel
Authors: Shokr A. M. , Shaat M. K.
Published: December 2013

This paper addresses the training and retraining programme for research reactor operating personnel. The factors that need to be considered in establishing an effective organizational structure for research reactor operation are discussed as they form the basis for the development of the initial training and retraining programme for the operating personnel. Different elements for a successful implementation of this programme such as training system, materials, and methods are also discussed. Application of these elements to the case of the Egypt Second Research Reactor (ETRR-2), which is a relatively large research reactor, is illustrated. The detailed content of the training programme is presented along with discussions that include the applicability of this programme to smaller research reactors.
Precursor analyses for German nuclear power plants
Authors: Babst S. , Gänssmantel G. , Stück R.
Published: December 2013

Precursor analysis is an internationally recognized method for quantifying the safety-relevance of operational events in nuclear power plants. Precursors are operational events in nuclear power plants which had no serious impact, but which could have led to serious impacts, if additional malfunctions had occurred. Examples of such operational events are component failures or transients, for example, the loss of main feedwater. On the basis of the probabilities for the occurrence of additional malfunctions or initiating events precursor analyses determine the probability with which these additional malfunctions during the event occurred would have led to core damage. This conditional probability is a measure for the safety relevance of the operational event occurred. Events, for which the probability of core damages is > 10–6, are internationally classified as “precursor”.
Early warning against airborne radioactivity in Bavaria: Measuring network for radioactive immissions
Authors: Faleschini J. , Mayer H. , Bielz M. , Hackl W. , Schulz T.
Published: December 2013

During the Chernobyl accident it became evident that there was no up to date information of the radiological situation in Bavaria as a whole. Therefore an online measuring network was implemented, which provides radioactivity data around the clock from all over Bavaria. It comprises measuring devices for gamma-dose rate, airborne aerosol and iodine activity and precipitation. The data are processed in accessory computers and transferred to the main network computers in Augsburg by means of a subcenter computer at the site of the IfR-station. Alarm levels are low in order to have an early warning against airborne radioactivity. New techniques for remote maintenance of the measuring devices and nuclide identification are being planned and tested.
Automatic calculation of sites for positioning of mobile gamma probes in case of an emergency at a Bavarian nuclear power plant
Authors: Faleschini J. , Falk H. , Gölling K.
Published: December 2013

In the early phase of a nuclear emergency many decisions have to be made by the authorities. So it is of great support when certain measures run automatically. Additional mobile gamma probes are therefore picked up by the local fire brigade at the nuclear power plant without further ado by the authorities. The gamma probes are then positioned at sites which are calculated automatically, taking into account the momentary weather situation. The sites are printed on a map handed over to the firemen. A location of a standby area after having set up the gamma probes is also given on this map.
Nuclear emergency preparedness and response in Germany
Authors: Miska H. , Bühling A.
Published: December 2013

Off-site nuclear emergency response in Germany is divided into disaster response under the responsibility of the Länder and measures for precautionary radiation protection pursuant to the Precautionary Radiation Protection Act under the lead of federal authorities. Early countermeasures at the regional level require a different management than long-term and comprehensive actions of precautionary radiation protection. As situations may arise in which measures of both approaches overlap with regard to place and time, it is essential to make thorough preparations in order to avoid problems with implementation.
Main results of the European project NURESIM on the CFD-modelling of two-phase Pressurized Thermal Shock (PTS)
Authors: Lucas D. , Bestion D. , Coste P. , Pouvreau J. , Morel Ch. , Martin A. , Boucker M. , Bodèle E. , Schmidtke M. , Scheuerer M. , Smith B. , Dhotre M. T. , Ničeno B. , Galassi M. C. , Mazzini D. , D'Auria F. , Bartosiewicz Y. , Seynhaeve J.-M. , Tiselj I. , Štrubelj L. , Ilvonen M. , Kyrki-Rajamäki R. , Tanskanen V. , Puustinen M. , Laine J.
Published: December 2013

The European Platform for NUclear REactor SIMulations, (NURESIM project 2005 – 2008) addressed the creation of a Common European Standard Software Platform for modelling, recording, and recovering computer data for nuclear reactors simulations. One work package of the project was dedicated to the analysis and improvement of CFD capabilities for the simulation of two-phase PTS problems. Some SB-LOCA scenarios lead to a situation in which the cold leg is partially or totally uncovered when the Emergency Core Cooling injection is activated. The resulting complex two phase flow can be divided in characteristic flow regions: the jet flow with a free surface between steam and water, the zone of jet impingement, the horizontal two-phase flow and the flow in the downcomer. Many phenomena have to be reflected in a simulation of each separate region, but also when the simulations are coupled reflecting the integral process which is required to predict the thermal loads at the RPV wall. After analyzing the experimental database available for CFD model development and validation and identifying shortcomings of the models different activities were dedicated to the simulation of single flow regions as well as the integral flow. Based on these experiences recommendations for the CFD-simulation of the two-phase PTS problem were obtained.
Investigation of coolant mixing in head parts of VVER-440 fuel assemblies with burnable poison
Authors: Tóth S. , Aszódi A.
Published: December 2013

Detailed CFD model for the head parts of the VVER-440 fuel assemblies with burnable poison has been developed. The coolant mixing was analyzed in some typical assemblies with this model and the signals of the in-core thermocouples above the selected assemblies were calculated. The investigations pointed out that the mixing is intensive in these assembly heads but the coolant is not perfectly mixed before reaching the thermocouples. Significant differences between the outlet average coolant temperatures and the thermocouple signals were revealed in the case of the fresh fuels. These deviations can cause about 6 % underestimations in the online monitored assembly powers unless a proper correction is introduced. The coolant mixing was also studied by means of numerical tracers and weight factors of selected rod bundle regions for the in-core thermocouple were determined. Using these weight factors and the outlet enthalpies of the assemblies' subchannels, the thermocouple signals can be corrected.
Algorithmic determination of fuel rod cladding burst time at elevated temperatures
Authors: Sauer G.
Published: December 2013

The integrity of fuel rod claddings during a loss of coolant accident (LOCA) is usually assessed by comparing calculated stresses or strains with the corresponding failure thresholds. Alternatively, the cladding integrity could also be evaluated on the basis of the time elapsed since the LOCA onset. To apply this assessment criterion the relationship between failure time and cladding loading must be available. The relationship can be established computationally by monitoring the evaluation of cladding inner and outer radius during the high temperature creep process. The cladding is classified as defective when the predicted inner radius equals or exceeds the anticipated outer radius. Using a suitable creep rate correlation it is shown which results are yielded by this procedure.
Accurate critical slab calculations for various degrees of anisotropy and for different reflection coefficients
Authors: Sjöstrand N. G.
Published: December 2013

In order to check new calculation results on critical slabs, numerical calculations have been done with a discrete ordinates method. The critical thickness of reflected slabs with various degrees of anisotropic scattering has been determined. In general, the agreement with the new calculation results is good. However, for thin slabs the deviations may be larger than one per cent, so some of these cases were studied in more detail.
Finite differences with exponential filtering in the calculation of reactivity
Authors: Díaz D. Suescún , Martinez A. Senra
Published: December 2013

A formulation for the calculation of reactivity using a recursive process is presented in this paper, as well as the treatment to reduce noise intensity that is found in the nuclear power signal. Using the history of nuclear power considered as the memory of such power and the filter exponentially adjusted with the least squares method, it is possible to reduce the nuclear power fluctuations without causing attenuation for the calculation of reactivity and with a smaller delay than that for low-pass filter of first order delay filter.
A discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium
Authors: de Abreu M. P.
Published: December 2013

In this article the transport model and inductive approach reported in a recent work have been extended, and a discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium is derived. At first, a mathematical description of the mixed beam transport problem that we are concerned with is given. The resulting mathematical problem is, according to Chandrasekhar, decomposed into uncollided and diffuse problems. After treating the uncollided problem analytically, a discrete ordinates formulation of the diffuse problem is considered. A set of spatially discretised equations that are free from spatial truncation error is then introduced. Finally, a discrete ordinates operator that replaces the entire multilayered medium in diffuse neutron transmission computations is derived inductively. To the best of our knowledge, such an operator has never been reported previously in neutron transport literature. Numerical results for a four-layer medium are presented to illustrate the computational merit of a computer code based on our diffuse transmission operator.
Efficiency of reflection coefficients in the isotropic neutron transport equation and computation of the critical-half thicknesses
Authors: Bülbül A. , Kara A. , Anlı F.
Published: December 2013

The critical slab problem of the reflected isotropic one-speed neutron transport equation is solved with the Chebyshev polynomial approximation. The efficiency of the reflection coefficient, R, in the neutron transport equation is obtained for different c values. For the solution, the Mark boundary condition is used. The values obtained from this approximation are compared with results obtained by the spherical harmonics method.
In-pile measurements of the fission gas pressure and comparison with ROFEM code results
Authors: Horhoianu G. , Ionescu D. V. , Pauna E. I.
Published: December 2013

Two experimental fuel elements with different UO2 grain size were irradiated in the C2 capsule in the TRIGA MT reactor of INR Pitesti to a burn-up of 178.9 MWh kg−1 U and linear powers between 45 to 60 kW m−1. Both elements were instrumented to measure internal gas pressure during irradiation. Experimental results presented in this paper represent data relative to two fuel elements with different UO2 grain size. The predictions of fuel performance code ROFEM in terms of internal gas pressure were compared with the experimental data. Experimental programs are in progress at INR Pitesti. New instrumented tests are planed to be performed in the C2 capsule in order to establish the pellet microstructure influence on fission gas release. This paper describes measuring techniques used at INR Pitesti and present experimental results of the in-pile tests comparatively with the ROFEM code predictions.
Nuclear model calculation on charged particle induced reactions to produce 85Sr for diagnostic and endotherapy
Authors: Kakavand T. , Sadeghi M. , Alipoor Z.
Published: December 2013

85Sr, having a half-life of 64.8 d, is an important radionuclide in nuclear medicine diagnostic and endotherapy. The ALICE/ASH code was used to calculate excitation functions for proton, alpha and deuteron induced on various targets that lead to the production of 85Sr radioisotopes using intermediate energy accelerators. Recommended thickness of the targets according to SRIM code was premeditated. The application of those data, particularly in the calculation of integral yields, is discussed and theoretical integral yields for different reactions were computed. To consider precision of ALICE/ASH code calculations, experimental excitation functions in several decay channels was compared with ALICE/ASH code data. The 85Rb(p, n)85Sr process was determined as most interesting one due to radionuclidic purity. The ALICE/ASH code predicts a maximum cross-section of about 798 mb at 11 MeV for this reaction.
Calculation of moderator circulation in IPHWR using a porosity approach
Authors: Goyal P. , Dutta A. , Singh R. K. , Ghosh A. K.
Published: December 2013

In the present configuration of the calandria for the 700 MWe Kakrapara Nuclear Power Plant, moderator inlet diffusers are directed upwards and the outlet is from the bottom of the calandria. Moderator circulation patterns and temperature distribution needs to be predicted to ensure adequate cooling margin for all channels. This study consists of two steps: at first, an optimized calculation scheme is obtained by comparison of the predicted results with the experimental data and by evaluating the fluid flow and temperature distribution. Then, in the second step, the analysis for the real 700 MWe IPHWR moderator under normal operating conditions has been performed with the optimized scheme. The present paper describes the methodology used for predicting the circulation pattern and temperature distribution in the moderator during normal operation using CFD code CFD-ACE+. The matrix of the calandria tubes in the core region is simplified to a porous media in which the momentum resistance model is used for pressure loss. The buoyancy effects due to internal heating and jet momentum effects through inlet nozzles have been considered in the analysis. The results show that the maximum temperature observed in the calandria is within the design limits during normal operation.
Sipping tests for the irradiated fuel elements of the TR-2 research reactor
Authors: Aytan Ö. , Büke T.
Published: December 2013

Sipping tests have been performed for fuel elements of the TR-2 reactor at Çekmece Nuclear Research and Training Center (ÇNRTC), in order to find out which one failed in the core. A sipping assembly has been constructed and placed in the pool of the TR-2 reactor. The assembly identifies leaking fuel elements by collecting and measuring137Cs that leak out from the defective fuel elements. 31 fuel elements in the reactor have been tested for the clad integrity. The measured137Cs activity of the fuel element with an identification number S-104 is a 10247 Bq/(0.3 l). This value is approximately 234 times greater than the average of the other tested fuel elements in the reactor.
CFD simulation of fibre material transport in a PWR core under loss of coolant conditions
Authors: Höhne T. , Grahn A. , Kliem S. , Weiß F.-P.
Published: December 2013

During a postulated cold leg LOCA with hot leg ECC injection, a limited amount of small fractions of the insulation material after passing the sump strainers can enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. The CFD simulations show that after starting the sump mode, the ECC water injected through the hot legs flows down into the core at so-called “brake through channels” located at the outer core region where the downward leg of the convection role had established. The hotter, lighter coolant rises in the center of the core. As a consequence, the insulation material is preferably deposited at the uppermost spacer grids positioned in the break through zones. This means that at the beginning the fibers are not uniformly deposited over the core cross section.
Generic experiments at the sump model “Zittauer Strömungswanne” (ZSW) for the behaviour of mineral wool in the sump and the reactor core
Authors: Alt S. , Hampel R. , Kästner W. , Kratzsch A. , Renger S. , Seeliger A. , Zacharias F. , Cartland-Glover G. , Grahn A. , Hoffmann W. , Krepper E. , Kryk H.
Published: December 2013

The investigation of insulation debris transport, sedimentation, penetration into the reactor core and head loss build up becomes important to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during loss of coolant accidents. Research projects are being performed in cooperation between the University of Applied Sciences Zittau/Görlitz and the Helmholtz-Zentrum Dresden-Rossendorf. The projects include experimental investigations of different processes and phenomena of insulation debris in coolant flow and the development of CFD models. Generic complex experiments serve for building up a data base for the validation of models for single effects and their coupling in CFD codes. This paper includes the description of the experimental facility for complex generic experiments (ZSW), an overview about experimental boundary conditions and results for upstream and down-stream phenomena as well as for the long-time behaviour due to corrosive processes.
Experiences with the coupled code system S3R/RELAP5-3D in training simulators
Authors: Belblidia L. A. , Höppner M.
Published: December 2013

The paper describes the implementation of S3R (core neutronics) and RELAP5-3D (RCS thermal-hydraulics) in a PWR training simulator for the Grohnde NPP located at the Kraftwerks-Simulator-Gesellschaft (KSG) in Essen, Germany. The models are briefly described as well as the coupling between these two codes and the interface with the rest of the simulator. The paper also describes the procedure that will be used to update the simulator core data after future core reloads. Results from the integrated simulator are presented.
Local heat transfer coefficient near the spacer grids
Authors: Cesna B. , Valincius M.
Published: December 2013

The paper presents the experimental data of the local heat transfer in the fuel rod assembly with honeycomb type spacer grids. It was established experimentally that the separation of the boundary layer behind the spacer grid increases the interchannel coefficient of transverse mixing of flow, and therefore, the local heat transfer. If the distance between the spacer grids is relatively small, they influence one another through the redistribution of mass flow in elementary channels. The local heat transfer behind the spacer grid varies with a distance from it. The local heat transfer can be determined within ±3% using similar equations for any elementary cell.
Cyclotron production of 85Sr by proton irradiation of natRb
Authors: Sadeghi M. , Alipoor Z. , Aslani G.
Published: December 2013

Excitation functions for protons induced on 85/natRb targets that lead to the production of 85Sr radioisotopes using TALYS 1.0 and ALICE/ASH codes were calculated. Rubidium chloride deposition on copper substrate was carried out via sedimentation method in order to produce Strontium-85. 520 mg RbCl, 208 mg ethyl cellulose (EC) and 4 mL acetone were used to prepare a layer of enriched rubidium chloride of 11.69 cm2 area and 62.2 mg · cm−2 thickness. The deposited target was irradiated at 20 A current and 15 MeV proton beam for 10 h and no degradation was observed. The 85Sr production yield was 1.4 MBq/mA · h. The target material was dissolved by 50 mL acetone. Dissolved Strontium-85 was separated from rubidium by Chelex 100 ion exchanger.
A detailed investigation of interactions within the shielding to HPGe detector response using MCNP code
Authors: Thanh T. T. , Tao C. V. , Loan T. T. H. , Nhon M. V. , Chuong H. D. , Au B. H.
Published: December 2013

The accuracy of the coincidence-summing corrections in gamma spectrometry depends on the total efficiency calibration that is hardly obtained over the whole energy as the required experimental conditions are not easily attained. Monte Carlo simulations using MCNP5 code was performed in order to estimate the affect of the shielding to total efficiency. The effect of HPGe response are also shown.
Advanced simulation and management software for nuclear emergency training and response
Authors: Rose K. W.
Published: December 2013

The importance of training of safety personnel to deal with real world scenarios is prevalent amongst nuclear emergency preparedness and response organizations. For the development of training tools we have committed to ensure that field procedures, data collection software and decision making tools be identical during training sessions as they would be during a real emergency. By identifying the importance of a fully integrated tool, we have developed a safety support system capable of both functioning in training mode and real mode, enabling emergency response organizations to train more efficiently and effectively. This new fully integrated emergency management tool is called S3-FAST also known as Safety Support Systems – Field Assessment Survey Tool.
Measurements required for evaluating a situation including radiological hazards
Authors: Prüßmann S.
Published: December 2013

For the evaluation of a situation including radiological hazards a wide variety of measuring devices and methods are available. By selecting the approach we have to decide whether

the results give us a reliable evaluation of the radiological hazards

the results are promptly achievable

qualified staff is available.

Hereby the dose burden of the response personnel and a forecast about the risks for the population and the environment have to be considered. In the training of the response personnel these aspects must be included.
Diffusion approximation for certain scattering parameters of the Anli-Güngör phase function
Authors: Öztürk H. , Anli F.
Published: December 2013

The first order approximations of the Legendre and Chebyshev polynomials are applied to neutron transport equation to calculate the diffusion length for one-speed neutrons in a uniform homogeneous slab. The Anli-Güngör (AG) phase function is used for the scattering kernel. Analytic and numerical results are obtained for the diffusion length of the neutrons in terms of scattering parameters of AG phase function and cross-section parameters. The numerical results for certain scattering parameters of AG phase function and selected values of cross-section parameter from first order approximations of both Legendre and Chebyshev polynomials are tabulated in the tables for comparison.