bluuebunny/crossref_metadata_2025_split huggingface dataset:
Application of theUNapproximation to the neutron transport equation in slab geometry
Authors: Bülbül A. , Ulutas M. , Anli F. 
Published: December 2013

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

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

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

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

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

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

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

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

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

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

Abstract
Increasingly wide application of Probabilistic Safety Assessment (PSA) to support a variety of issues dealing with operation and design of Nuclear Power Plants (NPP) imposes certain requirements on PSA quality in terms of its scope, level of detail, approaches, assumptions, and data used. If a unified set of balanced requirements is developed and consistently implemented, it would give confidence in using PSA results and insights in different applications and Risk-Informed Decision Making (RIDM); in addition, it would be possible to consistently compare PSA results for different plants. Such requirements are currently developed or being developed in several countries. The International Atomic Energy Agency (IAEA) is in the process of development of several publications in the IAEA Safety Standards Series that would provide guidance on the standard technical content of PSA. The paper discusses the latest developments at the IAEA in the area of PSA and provides a brief overview of the publications.
Determination and evaluation of the radiological situation using mobile measurements during the ground phase
Authors: Obrecht R. , Pohl H. , Schneider S. , Grimm Ch. , Neff U. , Coutinho P. , Müller U. , Mandel C. , Wilbois T. , Ren Y. , Chaves F. 
Published: December 2013

Abstract
The nuclear reactor remote monitoring system of the state of Baden-Württemberg (Kernreaktor-Fernüberwachung Baden-Württemberg – KFÜ BW) is implemented according to the recently renewed “recommendations for remote monitoring of nuclear power plants”. In Baden-Württemberg, the application area of the system covers both, the surveillance of internal procedures on one hand, and the handling of incidents or accidents on the other. The following paper shows the role of the KFÜ regarding the determination and evaluation of the radiological situation in the range of off-site emergency response. Progress is reported on the measurement conception and the technical possibilities for the investigation of the radiological situation after the end of the deposition of radio nuclides (ground phase).
The boundary problem in the 1D-CANDLE burn-up reactor
Authors: Seifritz W. 
Published: December 2013

Abstract
We have analytically solved the boundary problem for a solitary CANDLE-Burn Up wave in a slab reactor configuration using the affine approach which can be considered as the analogon to the Buckling-Concept for spatially/stationary critical zones in linear reactor physics. It sheds some light on the problems, we are confronted with, when we have to deal with nonlinear mathematics in this new field of reactor physics.
Nuclear level density parameters of some largely deformed nuclei of medical radionuclides in low-lying collective excitation modes
Authors: Okuducu Ş. , Aktı N. N. 
Published: December 2013

Abstract
In this study, the nuclear energy level density based on nuclear collective excitation mechanism has been identified in terms of low-lying collective level bands near the neutron binding energy. Nuclear level density parameters of some large deformed nuclei of radionuclides widely used in medical applications have been calculated by using different collective excitation modes of observed nuclear spectra. The calculated parameters have been used successfully in estimation of the neutron-capture cross section basic data for the production of new medical radionuclides. The investigated radionuclides have been considered in the region of mass number 150 < A < 200. The method used in the present work assumes equidistance spacing of the collective coupled state bands of the interest radionuclides. The calculated results have been compared with the compiled values from the literatures for s-wave neutron resonance data.
Targetry of Y2O3 on a copper substrate for the non-carrier-added 89Zr production via 89Y(p, n)89Zr reaction
Authors: Sadeghi M. , Kakavand T. , Taghilo M. 
Published: December 2013

Abstract
A thick layer of Yttrium oxide was deposited on cupper substrate via two special sedimentation methods for the production of 89Zr. For deposition of Y2O3 on cupper substrate, 435 mg of Y2O3, 152.25 mg of ethyl cellulose and 4 ml of acetone were used to prepare a Y2O3 layer of 11.69 cm2. Also, 435 mg of Y2O3, 174 mg of methyl cellulose and 7.25 ml of water were used. The targets were checked by SEM and thermal shock test. The deposited target was irradiated for 20 min at a current of 20 μA and a proton beam of 15 MeV.
Reflection on the ductility of irradiated zircaloy-4 fuel rod cladding
Authors: Sauer G. 
Published: December 2013

Abstract
An approach is presented that allows the computational determination of the ductility of irradiated zircaloy-4 cladding by transferring experimental results yielded under conditions atypical for normal power operation to typical loadings encountered in an usual pressurized water reactor. The transfer bases on a damage function that describes the gradual development of material separations during the cladding creep in the reactor. The numerical values of the needed equation parameters are derived from experimental results. Creep and damage evolutions are coupled leading to a simultaneous computation of both processes. The calculations indicate that the minimum strains that irradiated cladding material can accommodate are greater than 5% under pure pressure loading up to hoop stresses of 100 MPa and greater 7% when fuel swelling corresponding to a linear rod power of maximum 400 W/cm dictates the creep.
Cyclotron production of 101Pd/101mRh radionuclide generator for radioimmunotherapy
Authors: Enferadi M. , Sadeghi M. , Ensaf M. 
Published: December 2013

Abstract

101mRh is one of such radionuclides that has been considered as a potential candidate for targeted radiotherapeutic use, due to its nuclear decay and chemical properties. Electrodeposition of rhodium metal on a copper backing was performed in acidic sulphate. The target was bombarded with a current intensity of 120 μA (Ep = 29 → 25 MeV) for 30 min (60 μAh). Radiochemical methods were investigated to optimize the production of no-carrier-added 101Pd/101mRh. The use of a cyclotron target with radiochemical processes (i.e. electrodeposition, electrodissolution and ion-exchange column chromatography) were carried out to produce this radionuclide in high-specific activity and radiochemical form suitable for radiopharmaceutical syntheses.
Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction
Authors: Akimov P. , Obereisenbuchner L. 
Published: December 2013

Abstract
During a postulated loss-of-coolant accident, significant structural loadings on the reactor pressure vessel and its internals can be exerted due to the non-uniform and time-dependent pressure differences. The integrity of the internals must be secured, so as to ensure reactor shutdown and post-accident heat removal. Two developed computer codes to be discussed in this paper are used for the analysis of loads on the reactor pressure vessel internals. In the analyses the fluid-structure interaction effects are taken into account in the annular space between the pressure vessel and the core support barrel as well as in the upper plenum. This is a more realistic approach with respect to peak loadings and structural response frequencies.
Simulation of natural circulation in a rectangular loop using CFD code PHOENICS
Authors: Kumar M. , Borghain A. , Maheshwari N. K. , Vijayan P. K. 
Published: December 2013

Abstract
Single phase natural circulation in a rectangular loop is simulated using the PHOENICS code, a general purpose Computational Fluid Dynamics (CFD) code. The rectangular loop, having different operating power levels, has been modeled with the help of the Multiple Block Fine Grid Embedment (MBFGE) technique. The Co-located Co-variant Method (CCM) is used to simulate this loop in PHOENICS. Extensive experimental and CFD studies have been conducted on single phase natural circulation in a rectangular loop. The paper presents the results of three-dimensional CFD analysis for the prediction of steady state behavior in a rectangular loop and its comparison with experimental data. The results of code prediction and readily available experimental data show good agreement.
Analyses of loss-of-coolant accidents involving release of insulation material under consideration of the measures taken for PWR from expert's point of view
Authors: Huber J. , Becker M. , Unterrainer J. , Thiele T. , Reichel C. , Schiefer S. 
Published: December 2013

Abstract
The incident, which happened in the Swedish nuclear power plant Barsebäck 2 on 28th of July, 1992 brought about a re-evaluation of the design requirements for the control of loss-of-coolant accidents (LOCA) involving the release of insulation material that were valid in Germany up to that time. The current paper provides an overview on the aspects that need to be taken into consideration for analyses as well as a description of the important boundary conditions. In giving details of the analyses, the actual state is also described. It is also illustrated that sufficient examinations and tests have been performed for the German PWR plants and corresponding measures have been taken providing a sufficient basis for a final positive evaluation of the issue “LOCA involving release of insulation material”.
Analyses and estimation of insulation material release in E.ON-PWR under loss of coolant conditions
Authors: Koring R. 
Published: December 2013

Abstract
In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially and temporary clogged with mineral wool after steam jet induced releases of parts of the mineral wool insulation. Although Barsebäck NPP Unit 2 is a Boiling Water Reactor (BWR) this event induced large investigations to under-stand und maintain strainer clogging effects after a loss-of-coolant accident for all reactor types.
Especially for the German Pressurized Water Reactors (PWR) a program was launched by the German Utilities together with the plant manufacturer. The final estimations of all theoretical and experimental results caused different steps of modifications of the insulation material, the strainer area and mesh sizes in most of the German PWR including the E.ON PWR. Moreover, interactions and procedures were carried out such as back flushing actions to remove the mineral wool from the strainers supported by an additional differential pressure measurement at the strainers. All measures were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode and to improve the NPP safety.
Influence of nuclear data uncertainties on reactor core calculations
Authors: Klein M. , Gallner L. , Krzykacz-Hausmann B. , Pautz A. , Zwermann W. 
Published: December 2013

Abstract
To investigate the influence of nuclear data uncertainties on reactor core calculations systematically, the sampling based uncertainty and sensitivity software package SUSA developed at GRS was extended for the use with nuclear covariance data. Varied nuclear data are generated randomly corresponding to the uncertainty information from the covariance matrices. After performing a large number of calculations with these data, the results are statistically evaluated; this can be done not only for integral, but also for local output quantities like the assembly power distribution. The method is applied to multi-group Monte Carlo calculations stationary states of the PWR MOX/UO2 core transient benchmark, and to corresponding nodal diffusion calculations. Unexpectedly large uncertainties result for the radial power distribution. The uncertainties in the nodal results agree very well with those in the Monte Carlo reference results; thus, it is possible to apply the random sampling method to determine the influence of nuclear data uncertainties on transient core calculations.
BWR transient analysis with the coupled code system S3K/RELAP5
Authors: Jönsson C. , Moberg L. , Grandi G. , Judd J. 
Published: December 2013

Abstract
This paper describes the coupled code system S3K-RELAP5 and the application of the code system to BWR transients without scram. In addition short descriptions are given for application of the coupled system to fast transients and rod drop accidents. The models and methods of the code system are presented. Validation results are shown for a recorded transient event in the Forsmark 3 reactor. It is concluded that S3K-RELAP5 adequately captures the complicated interaction between physical processes in the reactor as well as the essential reactor protection and control systems, which qualifies it for applications to fast and slow transients, with or without scram, and Reactivity Initiated Accidents (RIA).
The development and assessment of TRACE model for Lungmen ABWR
Authors: Lin H.-T. , Wang J.-R. , Shih C. 
Published: December 2013

Abstract
For the nuclear power plant Lungmen (two blocks of Advanced Boiling Water Reactor (ABWR)) in Taiwan a plant model for the thermal hydraulic program TRACE (TRAC/RELAP Advanced Computational Engine) was developed. This model is and will be used for the simulation of normal and anomalous plant behaviour as well as for the simulation of incident scenarios. The presentation of this model is done in three steps: The first step is the development of a TRACE model of Lungmen nuclear power plant (NPP) which includes the vessel, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters were identified to refine the model further in the frame of a steady state analysis. The second step is the performance of TRACE transient analyses, such as MSIV closure direct scram (MSIVCD, MSIV = Main Steamline Isolation Valve) and loss of feedwater flow (LOFW). The above transient data of Final Safety Analysis Report (FSAR) are used to verify the Lungmen NPP TRACE model. The trends of their analysis results are roughly similar. It indicates that TRACE model is satisfying for the purpose of Lungmen NPP safety analyses. The third step is the prediction analysis of Lungmen NPP startup tests by using the TRACE model. The prediction analysis results of TRACE comply with the startup tests procedure criteria.