bluuebunny/crossref_metadata_2025_split huggingface dataset:
Nuclear emergency preparedness and response in Germany
Authors: Miska H. , Bühling A. 
Published: December 2013

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract

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

Abstract
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

AbstractSipping 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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.
Calculation of beta induced Bremsstrahlung exposure from therapeutic radionuclide 198Au in tissues, DNA and RNA
Authors: Jabal-Ameli H. , Sadjadi S. , Ahmadi S. J. , Sadeghi M. , Bakht M. K. 
Published: December 2013

Abstract
Gold-198 (βmax=0.96MeV (98.6%), γmax=0.412MeV (95.5%) and T1/2=2.7 days) is a well-known therapeutic beta emitter in the field of nuclear medicine, and is being used for the treatment of many different cancers. In the present study, the Bremsstrahlung exposure induced by 198Au in different human tissues, DNA and RNA has been calculated. The specific Bremsstrahlung constant (ΓBr), Probability of energy loss by beta during Bremsstrahlung emission (PBr) and Bremsstrahlung activity (Arelease)Br were estimated. We strongly recommend these parameters should be considered in absorbed dose calculations of radionuclide therapy via 198Au.
Condensation-induced water hammer in a horizontal pipe
Authors: Martin C. Samuel 
Published: December 2013

Abstract
Experiments were conducted for the purpose of investigating the phenomenon of condensation-induced waterhammer in an ammonia refrigeration system. Waterhammer was initiated by introducing hot ammonia gas over static subcooled ammonia liquid placed in a horizontal 146.3mm diameter carbon steel pipe approximately 6.0m long. By means of fast response piezoelectric pressure transducers and a high speed data acquisition system rapid dynamic pressures were recorded whenever a condensation-induced event occurred. Moreover, by means of top-mounted diaphragm pressure transducers sensing gas pressure the speed of liquid slugs propagating along the pipe was determined. Based upon careful interpretation of the results criteria for the transition from initial stratified flow to slug flow were established, similar to the so-called Taitel-Dukler correlation in two-phase flow literature.
New practice for the evaluation of rod efficiency measurement by rod drop at the NPP Paks
Authors: Pós I. , Parkó T. 
Published: December 2013

Abstract
Measurement of rod efficiency is an important part of nuclear reactor safety. In case of operating power reactors it is usual to measure all rod efficiency at beginning of cycle hot zero power condition. Measurement of all rod efficiency is technically complicated because the measured “dynamic” reactivity cannot be compared directly to calculated “static” value of control rod efficiency. At Paks Nuclear Power Plant a new method was established to solve this problem. A static computer code used for reload safety analysis was extended to calculate core kinetics as well. Using the developed code the rod drop measurement itself is simulated and the time dependence of flux is determined. In addition, using Monte Carlo calculations weight functions are determined to describe the detector signal response to neutron flux changes in the reactor core. Using this combined system the detector signal itself can be simulated to the actual cycle and condition. Thus the measurement directly can be compared to simulation and the result can be evaluated. The method is under introduction to NPP Paks measurement practice and the first results – which are shown in the paper, are very promising.