bluuebunny/crossref_metadata_2025_split huggingface dataset:
Wavelet techniques for the determination of the decay ratio in boiling water reactors
Authors: Sunde C. , Pázsit I. 
Published: December 2013

Abstract
The usefulness of wavelet transform and wavelet filtering techniques for the improvement of the estimation of the decay ratio (DR), characterising the stability of BWRs, is discussed. There are two distinct areas investigated. The first concerns the improvement of the quality of the traditionally used auto-correlation function (ACF) for the estimation of the DR, by trend elimination and denoising. The subsequent estimation of the DR itself is made by traditional methods such as the peak-to-peak method or curve fitting. The second area is the estimation of the DR by the use of continuous wavelet transform. The possibility of estimating two different DRs in case of dual oscillations, and in particular the finding of the higher DR, is also investigated. It was found that wavelet pre-processing does not always improve the estimation of the ACF of a non-ideal signal, compared to other methods; but for signals containing various trends and data scatter in the ACF, it brings a noticeable improvement. As an extension of the discrete wavelet methods, the continuous wavelet transform appears to be a promising candidate to determine the critical DR even in the case of two oscillations being co-existent with different stability properties. The methods investigated or developed here were also tested on measured data from Swedish BWRs.
Fire PSA of the French PWR 900 MWe series
Authors: Barrachin G. , Corenwinder F. , Vial E. 
Published: December 2013

Abstract
The Institute for Radioprotection and Nuclear Safety (IRSN) has developed a Fire Probabilistic Safety Assessment of the operating French 900 MWe PWR in order to organize into a hierarchy the items of equipments and the compartments, impacting the core melt frequency following a fire, and to reinforce deterministic safety analysis with functional analysis and realistic fire scenarios. The IRSN Fire PSA is composed of two stages: the first one concerns the selection of critical zones and the second one concerns the detailed study of the critical compartments, corresponding to the development of the fire scenarios and of the fire-induced core damage sequences. The first version of the IRSN Fire PSA has been updated by integrating the specificities of the Fire Action Plan, implemented by the utility on the French 900 MWe PWR. This study shows that four compartments are preponderant. The integration of the Fire Action Plan allowed decreasing the number of preponderant compartments. The integration of the symptom-oriented procedures and the removal of conservatisms should lead to a lower core damage frequency for these preponderant compartments. The results show that fire is not a negligible risk but the implementation of the Fire Action Plan has a significant positive impact on safety.
Modelling of Carbon-14, Iodine-129 and Cesium-137 releases from near surface radioactive waste disposal and their impact on environment and humans
Authors: Adliene D. , Skridaila N. , Motiejunas S. , Meurville C. 
Published: December 2013

Abstract
Near surface disposal of low and intermediate level radioactive waste is probably the best solution for a long term radioactive waste management today. For safety assurance all near surface disposal aspects and evolution scenarios should be reviewed and analyzed including repository design and site selection, because releases of radionuclides from the repository into the environment cause radiation exposure to the public. Mechanisms of 14C, 129I and 137Cs transportation from the repository into the environment, modelled for the hypothetical radioactive waste disposal facility in Lithuania, are investigated and the results of effective dose calculations for the main exposure pathways are discussed in this paper. The RESRAD-OFFSITE code has been used in this study for modelling purposes. It is shown that the radionuclide transport time into the environment, its radioactive contamination and the radiation exposure to the public mostly depends on the mobility and sorption properties of radionuclides and on the appropriate transport pathways from the repository into the environment.
Radiological consequences of potential sabotage attack to storage casks on the ISFSI site
Authors: Wang W. H. , Lin U. T. 
Published: December 2013

Abstract
Interim storage is one major program of fuel cycle. Because there are some individuals staying or living very close to the ISFSI site that is being constructed, this paper examines the potential radiological consequences of hypothesized events for storage casks. This study uses ORIGEN, HOTSPOT and RESRAD software to calculate the source terms, acute dose and long-term dose, separately. The results show that explosive-induced buoyancy would disperse the source much farther than the site boundary in a sabotage accident. Nevertheless, the amount of radioactivity released to offsite would be very low, and the maximum dose received by any individual staying or living in the vicinity of ISFSI, even under the most conservative scenario, would be very much less than the local public relocation dose limit.
AVR prototype pebble bed reactor: a safety re-evaluation of its operation and consequences for future reactors
Authors: Moormann R. 
Published: December 2013

Abstract
The AVR pebble bed reactor (46 MWth) was operated 1967 – 1988 at coolant outlet temperatures up to 990°C, which are suitable for process heat applications. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published but unresolved safety problems of AVR and draws conclusions for future pebble bed HTRs: Although the AVR was operated only for about 4 years at coolant temperatures > 900 °C its primary circuit is heavily contaminated with dust bound metallic fission products (90Sr, 137Cs), which create major problems in current dismantling. The end of life contamination reached several percent of a single core inventory. The AVR contamination was mainly caused by inadmissible high core temperatures, and not – as presumed in the past – by inadequate fuel quality only. The high core temperatures, which cannot be easily measured in pebble beds, are probably caused by the insufficiently examined pebble bed mechanics. Safety of future reactors requires a gas tight containment and in addition elaborate R&D or reduction of requirements concerning temperatures and burn-up.
Consideration of ageing within probabilistic safety assessment models and results
Authors: Čepin M. , Volkanovski A. 
Published: December 2013

Abstract
Ageing is a process, where the properties of systems and processes may degrade through the time and age. The objective of the paper is to analyze the possibilities of introduction of ageing directly into the probabilistic safety assessment. Theoretical models of ageing and examples of their application are investigated, which show, how the current models can be upgraded in sense to directly include the effects of ageing into the models of probabilistic safety assessment. This paper shows that consideration of ageing is theoretically a straightforward issue, which can be dealt with in several ways, which are demonstrated in the paper. The most important problem is the lack of data about the effects of ageing, which would suit to the detailed models of ageing. The results, which are obtained, do not show significantly different results compared to the initial models without direct consideration of ageing. The majority of risk measures change slightly or notably, but no major break through has been found. Contribution of ageing may be smaller then contribution of changes due to other parameters.
Atmospheric dispersion modelling applying advanced model approaches
Authors: Martens R. , Thielen H. 
Published: December 2013

Abstract
Incidents and accidents occurring in nuclear facilities, transport accidents or malevolent actions associated with radioactive material may involve a release of airborne radionuclides. For these situations, a realistic assessment of the radiological consequences for the public is of great importance with respect to adequate emergency measures. Within this context, the determination of the spatial distribution of the resulting airborne concentration as well as of the deposition of radionuclides on the ground, vegetation and building surfaces are necessary prerequisites. For such calculations, advanced atmospheric dispersion models are increasingly used. Their characteristics and fields of application are presented.
Legendre polynomial series of the infinite medium Green's Function in neutral particle transport theory: the half-space albedo problem
Authors: Güleçyüz M.Ç. , Kaşkaş A. 
Published: December 2013

Abstract
Series expansion of the infinite medium Green's function in terms of Legendre polynomials is used to solve the half-space albedo problem. The importance of this procedure is that it can be used for the solution of many different physical problems in plane geometry and can be extended to find the solution of the problems in other geometries.
Heat transfer to the building structures of the Ignalina NPP accident localisation system
Authors: Babilas E. , Rimkevičius S. , Urbonavičius E. 
Published: December 2013

Abstract
The performed thermal-hydraulic analysis of the Accident Localisation System (ALS) has shown, that in the case of a Maximum Design Basis Accident (MDBA), in the long term, a significant amount of energy (∼40%) is transferred to the building structures. Therefore, special attention should be paid to the correct simulation of the building structures and heat transfer processes. Free convection, forced convection, condensation and conductivity are the phenomena, which are connected with the heat transfer to the building structures and should be investigated. In this paper, the heat transfer models available in the COCOSYS code, which is applied for the analysis, are discussed. The influence of the heat transfer models on the behaviour of thermal-hydraulic parameters in ALS compartments during MDBA is investigated.
Production cross sections of medical 110,111In radionuclides
Authors: Büyükuslu H. , Kaplan A. , Yıldırım G. , Aydin A. , Tel E. , Bölükdemir M. H. 
Published: December 2013

Abstract
In this study, production routes of medical 110,111In radionuclides produced by 110Cd(p, n)110In, 111Cd(p, 2n)110In, 112Cd(p, 3n)110In, 107Ag(α, n)110In, 109Ag(α, 3n)110In, 111Cd(p, n)111In, 112Cd(p, 2n)111In, 113Cd(p, 3n)111In, 114Cd(p, 4n)111In, 109Ag(α, 2n)111In, 107Ag(α, γ)111In and 113In(n, 3n)111In reactions have been investigated up to 70 MeV incident particle energy. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the geometry dependent hybrid model and the cascade exciton model. The full exciton model has also been used to investigate the pre-equilibrium direct effects. Equilibrium effects are calculated according to the Weisskopf–Ewing model. The calculated results are compared with the experimental data taken from the literature.
Asymptotic solution of critical thickness with direct second kind Chebyshev polynomial approximation in forward and backward scattering
Authors: Yaşa F. 
Published: December 2013

Abstract
To develop a model with scattering medium, a consistent backward forward scattering kernel is described and applied to the transport equation in plane geometry. Asymptotic solutions of critical half thicknesses are computed by using the direct second kind Chebyshev polynomial method (UN).
Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors
Authors: Narkunas E. , Smaizys A. , Poskas P. , Kilda R. 
Published: December 2013

Abstract
Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit 1 reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases.
3D thermal hydraulic simulation of the hot channel of a typical material testing reactor under normal operation conditions
Authors: El-Morshedy S. El-Din , Salama A. 
Published: December 2013

Abstract
The hot channel in a typical Material Testing Reactor (MTR) is subjected to 3D simulation. Because of the existence of similarity planes, only a quarter of the hot channel including meat thickness, clad, and coolant channel is considered for CFD analysis using the FLUENT code. For the simulation, steady state normal operation regime at the reactor nominal power is assumed. In order to build confidence in our modeling approach, the results obtained in this work are compared with those obtained from the one-dimensional simulation code, MTRTHA. That is, modified variables were generated in order to match those obtained by MTRTHA and to allow comparisons. Quite good agreement is generally observed, however, the maximum clad surface temperature predicted by the 3D calculations, located at the clad mid-width, is higher than the 1D prediction by about 8°C but still below the onset of subcooled boiling by adequate safety margin. The results show quite interesting 3D patterns in both the flow field and the heat transfer. Temperature profiles, velocity profiles and contours are all presented to highlight the essential 3D features of this system.
Effect of thermal gap conductance for MoO3 ampoules irradiated in a high neutron flux
Authors: Abdelhady A. 
Published: December 2013

Abstract
During the irradiation process of MoO3 powder with in a high neutron flux, energy deposited in the powder must be released to avoid energy accumulation. The temperature increasing in the powder due to energy accumulation may cause powder phase change and under certain conditions the temperature may reach the melting point. An analytical model was developed to study the effect of gap conductance on the temperature distribution in the powder during the irradiation process. The gap conductance model was studied for helium and nitrogen gases at different pressures to obtain the optimum conditions verifying the safety in heat transfer point of view. It was found that the gap conductance is a function of gas pressure. The model provided the optimum conductivity for the gap filled with helium gas at 1 atm pressure.
U1 approximation to the neutron transport equation and calculation of the asymptotic relaxation length
Authors: Öztürk H. , Bülbül A. , Kara A. 
Published: December 2013

Abstract
The U1 approximation is used to determine the asymptotic relaxation length (diffusion length) for one-speed neutrons in a homogeneous slab. The method is based on the series expansion of the neutron angular flux in terms of the Chebyshev polynomials of second kind and then calculating the diffusion length by applying the first order approximation to transport equation. Analytic and numerical results are obtained for the diffusion length and compared with the ones obtained from the method of separation of variables and simple diffusion theory (P1 approximation).
Determination of effects of burn up and reflector material on the kinetic parameters for open pool reactor using MCNP code
Authors: Shawer M. , Aly N. , Nagy M. , Amin E. 
Published: December 2013

Abstract
The analysis of research reactor transients depends on the kinetic parameters (i.e. effective neutron generation time Λ∗ and the effective delayed neutron fraction βeff).
In this work we will calculate these parameters using the MCNP code. Effects of burn up and of Be reflector and cobalt irradiation device on kinetic parameters were also studied using the MCNP code. The results were compared with reported values. The calculations were performed for the start-up core configuration (reference core configuration) at different burn up steps with equilibrium concentration and with zero concentration of Xe, and repeated for the sixth core configuration (core configuration during the study).
CFD analyses of fibre transport and fibre deposition at plunging jet conditions
Authors: Krepper E. , Cartland-Glover G. , Grahn A. , Weiß F.-P. , Alt S. , Kratzsch A. , Renger S. , Kästner W. 
Published: December 2013

Abstract
The investigation of insulation debris generation, transport and sedimentation becomes important with regard to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during all types of loss of coolant accidents. A joint research project on such questions is being performed in cooperation between the University of Applied Sciences Zittau/Görlitz (HSZG) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). The project deals with the experimental investigation of particle transport phenomena in coolant flow and the development of CFD models for its description (see [10–12]). While the experiments are performed at the University at Zittau/Görlitz, the theoretical modelling efforts are concentrated in Rossendorf. In the current paper, the basic concepts for CFD modelling are described and feasibility studies are presented. The model capabilities are demonstrated via complex flow situations, where a plunging jet agitates insulation debris.
Determination of natural uranium, thorium and radium isotopes in water and soil samples by alpha spectroscopy
Authors: Hao Le Cong , Van Tao Chau , Van Dong Nguyen , Van Thong Luong , Linh Duong Mong 
Published: December 2013

Abstract
In this study, a simple procedure for the determination of natural uranium, thorium and radium isotopes in water and soil samples by alpha spectroscopy is described. This procedure allows a sequential extraction polonium, uranium, thorium and radium radionuclides from the same sample in two to three days. It was tested and validated with the analysis of certified reference materials from the IAEA.
Lessons learnt from PSA for new and advanced reactors in Russia
Authors: Tokmachev G. , Morozov V. 
Published: December 2013

Abstract
Customer requirements to probabilistic safety targets are usually stronger than existing Regulatory or IAEA ones. It appears that industry takes the lead over regulation in this case and forces the designer to find and implement appropriate means to enhance safety, which sometimes have no reference to practical experience. On the other hand, regulatory documents and the existing PSA methodology are mainly oriented to operating plants. This creates problems when developing a PSA as well as performing regulatory reviews. The scope of the PSA may be different depending on a design stage such as the development conceptual, basic or detailed design. In addition, the base case PSA is usually performed for NPP in design. However, a customer may require additional PSA applications to consider, for instance, risk monitoring. In this case the scope of the PSA should be extended to implement special attributes of the application needed that often requires specific information not available at the design stage. Lack of design information affecting PSA development may be associated with incompleteness of the design that is typical for interim design stages and communication problems caused by the involvement of many different companies in the deign activity. To deal with this issue bounding technologies and the iterative PSA development are used. However this sometimes contradicts to the “best estimate” approach recommended by regulatory guides. PSA development for advanced NPPs has raised some issues originated from unknown new components, processes and technologies incorporated into the design of an advanced plant. The paper addresses some issues resolved while carrying out PSAs for advanced NPPs. Some PSA results for new advanced VVER plants under construction and the first lessons learnt from the Fukushima accident are also discussed.
Cosmic ray angular distribution employing plastic scintillation detectors and flash-ADC/FPGA-based readout systems
Authors: Hai V. H. , Dao N. Q. , Nomachi M. 
Published: December 2013

Abstract
It is known that secondary cosmic rays are high energetic particles which are products of shower particles, when primary cosmic rays from outer space hit the atmosphere molecules. Many studies of cosmic rays show that cosmic flux depends on the depth of atmosphere. At ground level, most secondary cosmic rays are muons, a type of charged particle, and have angular dependence. The purpose of this article was to develop the telescope with two plastic scintillation detectors to investigate the angular distribution of cosmic rays at ground level. Directions of investigation were carried out vertical, 45-degree oblique and horizontal directions with respect to earth surface, approximately from North to South. Electronic readout system was developed from Flash Analog Digital Converter (Flash-ADC) of 8 bits-250Ms/sec, and Embedded Field Programmable Gate Array (FPGA)-based trigger. LabVIEWTM interface was written for controlling trigger system and taking data. For each direction measurement, the deposited energy spectra in the scintillators were obtained. The interest of cosmic-ray events was analyzed for counting. Angular distribution was obtained quantitatively. The experiment has been done at University of Science-HCMC.
Studies on boiling water reactor design with reduced moderation and analysis of reactivity accidents using the code DYN3D-MG
Authors: Rohde U. , Pivovarov V. , Matveev Y. 
Published: December 2013

Abstract
The multi-group version of the DYN3D reactor dynamics code was used for calculations for a new concept of a boiling water reactor with tight lattice of fuel rods and reduced neutron moderation. For that purpose, a 5-group cross section library was prepared and connected to the DYN3D code. Comparison calculations with the steady-state finite-difference code ACADEM showed a very good agreement. The capability of the DYN3D multi-group code in modeling transients in boiling water reactors with tight fuel element lattices was demonstrated by the analysis of two reactivity accidents initiated by the ejection of one control rod and unauthorized withdrawal of a control rod bank from the reactor core. The corresponding analyses were performed for begin of cycle conditions, when the considered control rods are at their maximum insertion depth.
Dose contributions due to radiation scattered by air (skyshine) in the case of x-ray machines
Authors: Sahre P. , Kaden M. , Schönmuth T. , Pawelke J. , Naumann B. , Reichelt U. 
Published: December 2013

Abstract
Radiation transport simulations had to be done in preparation of operation of the X-ray tube ISOVOLT 320 kV/13mA in a special laboratory. At first simulation was done without shielding the roof of the laboratory, showing a dose rate maximum of more than 100 mSv/h. This dose rate results in a skyshine dose rate of at most 2 μSv/h in the surrounding of the building without shielding the roof. For similar geometries the skyshine is negligible for dose rates at the unshielded roof of less than 3 mSv/h (exclusion area).