bluuebunny/crossref_metadata_2025_split huggingface dataset:
PSA Level 2 – An optimized methodology
Authors: Torri A. 
Published: December 2013

Abstract
A methodology for a Level 2 PSA is described that provides a more efficient structure than the two methods recommended in the German Leitfaden. This is accomplished by merging the top events for the system and accident management functions from the Level 2 space, referred to as the containment system event tree (CSET) as a direct extension of the Level 1 model and by introducing the containment phenomenological event tree (CPET) to represent only the phenomenological Level 2 top events. The interface between the Level 1+ CSET and the CPET is defined as the plant damage states (PDS) with the focused objective to minimize the effects of sequence to sequence variability on the CPET failure fractions. A clear and consequent distinction is made between the statistically random events modeled in the Level 1+ CSET and the deterministic physical processes with large state of knowledge uncertainties modeled in the CPET. It is shown that as a logical consequence the resulting CPET failure fractions are numbers and not distributions, and these numbers are consistent with the interpretation as the degree of confidence probabilities for the outcome of uncertain physical processes. While both methods recommended in the Leitfaden will yield correct results if properly applied, the proposed method will result in a more streamlined model that will be easier to trace, understand and review, and it will result in lower release category uncertainties by minimizing the sequence to sequence variability through the elimination of the intermediate pinch points and through the focused use of the PDS.
Analysis of external flooding and tsunamis for nuclear power plants at tidal rivers
Authors: Berg H.-P. , Winter C. 
Published: December 2013

Abstract
The German regulatory body has issued technical documents on methods and data for probabilistic safety assessment, elaborated for a comprehensive integrated safety review of all nuclear power plants in operation. These documents contain, among others, a graded approach for the probabilistic assessment of external flooding. Main aspects are explained such as the underlying probabilistic considerations and the mathematical procedures for the calculation of exceedance frequencies for nuclear power plants at tidal rivers in Germany. Morever, extreme events such as tsunami waves have been investigated.
The Hamburg polonium-210 case: Monitoring of internal contaminations in a malevolent act
Authors: Dalheimer A. , König K. , Noßke D. , Schäfer I. 
Published: December 2013

Abstract
The dead of Alexander Litvinenko in London in the late 2006 was linked with the visit of Dimitri Kovtun. So an investigation was started by the Hamburg Police. As the presence of polonium-210 was uncertain, the action was supported by the Federal Defense against Nuclear Hazards (NGA) of the Federal Office for Radiation Protection (BfS). Radiation measurements in some flats in Hamburg confirmed the suspicion of polonium-210 contaminations. As incorporations by ingestion or inhalation could not generally be excluded, the Coordinating Office on Incorporation Monitoring of the BfS (Leitstelle Inkorporationsüberwachung des BfS) was requested to clarify this issue. The BfS office then coordinated the required bioassay analyses of urine samples carried out at four officially appointed in-vitro laboratories for incorporation monitoring. The group of people to be examined included the members of the family involved living in the contaminated flats, the emergency forces, as well as potentially exposed members of the public. The calculation of the internal dose was based on standard assumptions and also conducted by the BfS. The internal radiation doses proved to be within the range of radiation doses due to natural exposures. Retrospectively it should be stressed that the collaboration and cooperation of all institutions involved in the operation in Hamburg was extremely good despite their differing organisational structures. As a result of the incident in Hamburg various middle and long-term measures have been taken since then by the BfS Coordinating Office on Incorporation Monitoring.
Assessment of radiological consequences due to routine atmospheric discharges from nuclear power plants
Authors: Kocar C. , Sökmen C. Niyazi 
Published: December 2013

Abstract
The radiological consequences of operational atmospheric discharges from nuclear power plants are analyzed by using coupled long-range atmospheric dispersion models with time-dependent dose and risk assessment models. The coupled model is applied to atmospheric discharges from twelve assumed nuclear power plants located at three different sites in Turkey. Individual as well as collective doses and risks are analyzed. The effects of multi-location food supply and interregional movements of people in the computational domain are also investigated while computing individual effective dose and risk values. The values computed are found to be negligible compared with values from natural background and exposures from nuclear weapon testing.
Some affine solutions for CANDLE burn-up waves in 1D-geometry
Authors: Seifritz W. 
Published: December 2013

Abstract
Using the method of “affine solutions” three interesting phenomena in a CANDLE Burn-Up strategy were studied and were cast analytically into mathematical formlulae: a) the problem of two burn-up waves one behind the other (“enslavement principle”) b) the problem of the frontal collision of two burn-up waves (“Scattering”), and c) the problem of two identical burn-wave twins moving apart from each other after central ignition (“Breathing Out Phenomenon”).
To the knowledge of the author this is the first time where such affine problems of the CANDLE Burn-Up wave physics were solved.
Simulation of H2 distribution and combustion in LWR containments using Lumped Parameter Codes
Authors: Burkhardt J. , Klöcker M. , Brähler T. , Koch M. K. 
Published: December 2013

Abstract
This article focuses on the validation of the Lumped Parameter (LP) computer codes COCOSYS and ASTEC in the field of containment phenomena during severe accidents. The release of hydrogen (H2) as an open jet from the primary circuit into the containment might cause atmospheric H2-stratifications which in turn may result in an ignitable atmospheric mixture. If this mixture is ignited, the integrity of the containment can be endangered due to the pressure and temperature rise, resulting from the combustion process. The emphasis of the paper lies on these three phenomena referring to possible successive phases of a loss of coolant accident.
CFD modelling of insulation debris transport phenomena in water flow
Authors: Krepper E. , Cartland-Glover G. , Grahn A. 
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 and the Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow and the development of CFD models for its description. While the experiments are performed at the University at Zittau/Görlitz, the theoretical modelling efforts are concentrated at Forschungszentrum Dresden-Rossendorf. In the current paper the basic concepts for CFD modelling are described and feasibility studies are presented.
Closed loop auto control system software for Miniature Neutron Source Reactors (MNSRs)
Authors: Iqbal M. , Qadir J. , Bhatti T. K. , Abbas Q. , Mirza S. M. 
Published: December 2013

Abstract
A closed loop auto control system has been developed and implemented at the Pakistan Research Reactor-2 (PARR-2). For interfacing of the site signals with a computer, a data acquisition card was employed. The system utilizes nine signals and, on their basis, carries out continuous on-line analysis, maintaining safe operation. On-line reactivity measurement is included in this software. Also, it generates control signals for keeping the reactor at the desired power level in auto-mode round-the-clock without human intervention. In case of abnormal conditions for either power or temperature or radiation level, alarms are initiated and if their levels reach beyond prescribed safe operation regime, automatic shutdown sequence is initiated. The control system has been thoroughly tested for various postulated scenarios. The test results have been in good agreement with the expected response.
Chebyshev polynomials expansion method for solving the one-dimensional transport equation in spherical geometry
Authors: Yaşa F. 
Published: December 2013

Abstract
Using certain well-known properties of Chebyshev polynomials, an efficient analytical approach to evaluate the Boltzmann equation is presented in one-dimensional spherical geometry. The procedure is based on the expansion of the angular flux in terms of a series of Chebyshev polynomials.
Development of a decommissioning strategy for the MR research reactor
Authors: Bylkin B. , Craig D. , Fecitt L. , Gorlinsky Y. , Harman N. , Jackson R. , Kolyadin V. , Lobach Y. , Pavlenko V. 
Published: December 2013

Abstract
A description of the selected decommissioning strategy for the research reactor MR at the site of the Kurchatov Institute in Moscow is presented. The MR reactor hall is planned to be used as a temporary fuel store for the other research reactors on the site. On the basis of the site-specific conditions and overall decommissioning goals, it was identified that phased immediate decommissioning is the preferable option. The current status of the reactor, expected final conditions and the sequence of decommissioning works are shown.
Methods for thin foil thickness determination by using alpha spectroscopy
Authors: Hao L. C. , Nhon M. V. , Tao C. V. 
Published: December 2013

Abstract
In both, industry and scientific research, there is a demand for the determination of material thicknesses. In this work, two methods are described which are very simple, fast, effective and low-cost to determine the exactly thickness of thin foils (μm). Both methods are based on the energy loss of alpha particles when passing through the foil and the detection by high energy resolution alpha spectroscopy. The strengths and weakness of the two methods are discussed.
The time-dependent P1 model for the neutronics of multiplying systems: a review
Authors: Coppa G. G. M. , Dulla S. , Ravetto P. 
Published: December 2013

Abstract
The derivation of the time-dependent P1 model by the spherical harmonics expansion approach to solve the neutron transport equation is reviewed and its main features and physical characteristics are illustrated. The telegrapher's equation is presented and its formulation for multiplying systems is discussed. In particular, it is evidenced that the presence of a time derivative of the source terms in the telegrapher's equation implies the appearance of time derivatives of the flux and of the precursor concentrations. The role of these terms is evaluated and discussed.
The albedo problem for pure-quadratic anisotropic scattering with İnönü scattering
Authors: Türeci R. G. 
Published: December 2013

Abstract
In this study the half-space albedo problem has been investigated for highly anisotropic scattering. For this highly anisotropic scattering the İnönü scattering has been added to the pure-quadratic anisotropic scattering. The general solution of the one-speed neutron transport equation can be written as linear combination of the solutions which involve these scatterings. Hence, Case's eigenfunctions and the related semi-analytical methods can be used to get the albedo calculations. Some selected numerical data have been tabulated. To compare these results with the İnönü scattering values the linear interpolation technique has been used. Successful results have been achieved.
Pertinence of tests used for conformity assessments of fire resistance of nuclear power plant components – opinion of several institutions in Europe
Authors: Bruynooghe C. , Röwekamp M. , Hostikka S. , Richard J. , Maillet E. , Neugebauer W. , Zdeněk T. 
Published: December 2013

Abstract
Fire resistance of structures, systems and components (SSC) is widely identified as a fundamental issue for fire risk assessment at nuclear power plants (NPP). It is currently supported by several national or international ambitious experimental programs. Fire risk assessment at NPPs is performed under a series of hypothesis, e.g. the fire load. A problem will arise if tests used for conformity assessment do not appropriately represent the hypothesis selected during the fire risk analysis. In this case, SSCs may possibly be not meeting the expectations related to their fire resistance capacity supposed by the fire risk analysis. Ways of tackling this issue involve a regulatory approach and adapted requirements for components. While the widespread regulatory approach consists in seeking enveloping scenarios and conservative assumptions, a better knowledge of SSCs behaviour under for NPPs typical situations may be also desirable. The Institute for Energy of the European Commission's Joint Research Centre (JRC) at Petten (NL) launched a questionnaire with the aim of gathering opinions of regulatory bodies and their technical support organisations and manufacturers in Europe on this topic. Major goal was to estimate the relevance of the issue according to the perception of the specialists. The present paper collects the answers given by seven entities (AREVA, GRS, iBMB, IRSN, SJUB, SUEZ-TRACTEBEL, and VTT) and makes suggestions for addressing the issues identified by the participants.
Analysis of the processes in the target cooling system of the W7-X fusion experiment
Authors: Kaliatka T. , Kaliatka A. , Kačegavičius T. , Naujoks D. 
Published: December 2013

Abstract
Nuclear fusion has the potential to cover a major part of mankind's energy needs under the conditions of ever-increasing population and energy consumption. Wendelstein 7-X (W7-X) is a fusion experiment currently being built in Greifswald, which shall demonstrate the capability of the stellarator concept on the way to a fusion power plant. This paper presents analyses of processes in the target cooling system with the codes RELAP5 and ASTEC. The loss-of-coolant accident (LOCA) during the “baking” operation mode of W7-X (i.e. under the most critical conditions because of the high temperature and pressure applied) is analysed. The response of the W7-X “baking” and main cooling circuits and the amount of coolant discharged into the plasma vessel are investigated.
Performance analysis of 233U for fixed bed nuclear reactors
Authors: Şahin S. , Acır A. , Şahin H. M. 
Published: December 2013

Abstract
Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233UO2/ThO2 as an alternative to low enriched 235UO2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∼2% 233UO2 with the mixed 233UO2/ThO2 fuel. At higher 233U fractions, reactor criticality rises rapidly and exceeds keff > 1.5 already by 9% 233UO2. With 100% 233UO2, start up criticality can reach keff = 2.0975. Time dependent reactor criticality keff and fuel burn up have been investigated for two different mixed fuel 233UO2/ThO2 compositions, namely: 4% 233UO2 + 96% ThO2 for a reactor power of 40 MWel (120 MWth) and 9% 233UO2 + 91% ThO2 for a reactor power of 70 MWel (210 MWth). Sufficient reactor criticality (keff > 1.06) for continuous operation without fuel change can be sustained during ∼5 and 12 years with 4% and 9% 233UO2 fractions in the mixed fuel, leading to burn ups of ∼36000 and >105000MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible.
Application of the Chebyshev polynomial (T

N
and U

N
) approximation to reflected slab geometry in the neutron transport equation and computation of critical half thicknesses
Authors: Bülbül A. , Anlı F. 
Published: December 2013

Abstract
The transport of one-speed neutrons has been studied in isotropically scattering slab using ultraspherical polynomial approximation whose sub-cases are spherical harmonics approximation and Chebyshev polynomial approximation of first and second kind with reflective boundaries. An integral equation for the total flux has been derived and solved numerically for different values of c and reflection coefficients. In the solution, Marshak type vacuum boundary conditions were used. The critical thicknesses obtained by the Chebyshev polynomial method are in good agreement with literature values. Computations were made by using the TN, UN and PN approximations for comparison.
Analysis of fuel rod behaviour during limiting RIA in RBMK plants
Authors: Pabarcius R. , Kaliatka A. , Marao A. 
Published: December 2013

Abstract
Typically, reactivity initiated accidents for nuclear power plants with RBMK type reactors are classified as design basis accidents. However, for assessment of the real probability of these events, such kind of incidents should be considered more seriously and the same acceptance criteria as for operational occurrences should be applied to them. The spurious withdrawal of control rods, as limiting reactivity initiated accident (RIA), was analysed and results of the study are presented in this paper. The power distribution in fuel channels located around the affected control rods was calculated using the QUABOX/CUBBOX-HYCA code. Evaluating possible uncertainties it was found that linear power, i. e. heat generated per unit length of coolant channel, exceeds the acceptance criterion by a factor of 1.21. This paper presents the discussion how such a peak of power influences the behaviour of RBMK-1500 reactor fuel rods. For this purpose, the FEMAXI-6 code was applied. Results of the analysis show, that after increasing and subsequently decreasing of the linear power during an accident most parameters (pressure inside fuel rods, radial displacements, fuel clad gaps, etc.) return to their initial values and no phenomena occurred that would preclude the resumption of normal operation after termination of such an event.
Comparison between CAREB code calculations and LOCA test results in the FUMEX III project
Authors: Horhoianu G. , Ionescu D. V. , Pauna E. I. 
Published: December 2013

Abstract
The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO2 on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation.
Review and investigations of oscillatory flow behaviour of a horizontal ceiling opening for nuclear containment and fire safety analysis
Authors: Sharma P. K. , Singh R. K. , Ghosh A. K. 
Published: December 2013

Abstract
In the thermal hydraulics codes developed for fire safety analysis and for containment thermal hydraulic analysis, junctions in the multi-compartment geometries is often modeled as uni-directional junctions. However, ceiling junctions are known to depict unstable/oscillatory bi-directional flow behavior. Detailed investigations have been carried out to understand the unstable flow behaviour of a junction by analyzing an earlier reported experiment and its subsequent two dimensional numerical RANS based study of fire in an enclosure. The authors attempt more realistic and desired three dimensional and inherently transient large eddy simulations using a computer code Fire Dynamics Simulator (FDS). The paper presents the details of the analysis, the results obtained and further studies required to be conducted so that the findings can be applied to the fire/containment thermal hydraulics analysis codes successfully.
Neutron multiplication in source driven subcritical nuclear systems
Authors: Göksu A. , Geçkinli M. 
Published: December 2013

Abstract
In this work the neutron multiplication in a source driven subcritical nuclear system is studied for a particular physical model of symmetrical slab geometry. On both sides of a central spallation source region, bare fissile blankets are placed. The neutronics of the system is represented in terms of two-group diffusion equations. We elaborate on the development and evaluation of the multiplication factors for source and fission neutrons (and). The results are compared with the conventional multiplication factor of the fissile system. In the present static context, is important only from the safety point of view, whereas the other two factors are figures of merit measuring the effectiveness of a source neutron to start a fission chain and that of any one fission neutron to maintain a fading chain which further contributes to source multiplication.
Analysis of reactor lattice and core parameters in view of nuclear data modifications
Authors: Becker M. 
Published: December 2013

Abstract
A detailed assessment of the JEFF 3.1 and ENDF/B VII.0 nuclear data libraries with the KANEXT code system has been performed on each level of the reactor simulation steps.
This allowed for the sensitivity estimation of the lattice and core parameters to changes in nuclide and specific reaction cross section. The direct calculation of cross section data influence, in particular for specific reaction types, is possible due to the unique substitution feature of the group constant module GRUCAL.
Further on, the different impact of several reaction types on the lattice versus core calculation points out the “engineering uncertainty” using the two major libraries for the identical problem.
MCNP(X) and KANEXT cell models can be mutually adjusted such that differences between stochastic and deterministic data processing can be identified.
The current analysis emphasizes, that the two major nuclear data libraries JEFF 3.1 and ENDF/B VII.0 exhibit to some extent compensation of deviations rather than reliable integral data calculation.
On a special type of a fast reactor application we show that exchange of JEFF 3.1 and ENDF/B VII.0 cross sections introduce large discrepancies. A major part is attributed to the inelastic cross section. Further investigated discrepancies concerning the capture and the fission cross section in the epithermal energy range might be also a concern for LWR which was beyond the scope of the current study.