bluuebunny/crossref_metadata_2025 huggingface dataset:
A model for calculation of forward isotropic scattering with application to transport equation in slab geometry
Authors: Yaşa F. , Anli F.
Published: December 2013

In order to develop a general procedure, the scattering kernel is described in a consistent manner. By using this scattering kernel, eigenvalues and eigenfunctions and critical slab thicknesses are computed with a conventional spherical harmonics method (PN). The basis of this method relates to the generating function of Legendre polynomials. This article has improved analytical methods and numerical values. Numerical results indicate that the critical thickness varies with forward isotropic scattering. Results are discussed and compared with isotropic and anisotropic results using various methods available in the literature. Eigenvalues and critical thicknesses are calculated using the new phase function instead of conventional expansions.
Heuristic geometric “eigenvalue universality” in a one-dimensional neutron transport problem with anisotropic scattering
Authors: Gonçalves G. A. , de Vilhena M. T. , Bodmann B. E. J.
Published: December 2013

In the present work we propose a heuristic construction of a transport equation for neutrons with anisotropic scattering considering only the radial cylinder dimension. The eigenvalues of the solutions of the equation correspond to the positive values for the one dimensional case. The central idea of the procedure is the application of the SN method for the discretisation of the angular variable followed by the application of the zero order Hankel transformation. The basis the construction of the scattering terms in form of an integro-differential equation for stationary transport resides in the hypothesis that the eigenvalues that compose the elementary solutions are independent of geometry for a homogeneous medium. We compare the solutions for the cartesian one dimensional problem for an infinite cylinder with azimuthal symmetry and linear anisotropic scattering for two cases.
European clearinghouse on nuclear power plants operational experience feedback
Authors: Ranguelova V. , Bruynooghe C. , Noël M.
Published: December 2013

Learning from operational experience and applying this knowledge promptly and intelligently is one of the ways to improve the safety of Nuclear Power Plant (NPP). Recent reviews of the effectiveness of Operational Experience Feedback (OEF) systems have pointed to the need for further improvement, with importance being placed on tailoring the information to the needs of the regulators. In 2007, at the request of a number of nuclear safety regulatory authorities in Europe, the Institute for Energy of the European Commission's Joint Research Centre (EC JRC) initiated a project on Nuclear Power Plant operational experience feedback, which adopts an integrated approach to the research needed to strengthen the European capabilities for assessment of NPP operational events and to promote the development of tools and mechanisms for the improved application of the lessons learned. Consequently, a so-called “European Clearinghouse” on NPP OEF was established, which includes scientific officers from the EC JRC, a number of European nuclear safety regulatory authorities and some of their Technical Support Organizations (TSOs). The paper discusses the activities implemented in 2008 within the framework of the European Clearinghouse on NPP OEF (hereinafter called the European NPP Clearinghouse) and provides an overview of the main conclusions drawn from the safety studies performed. Outlook of the activities carried out in 2009 are given.
Steady state thermal hydraulic calculations for MTR plate-type research reactors
Authors: Abdelrazek I. D. , Zidan W. I. , Shokr A. M. , Gaheen M. A.
Published: December 2013

A computation code THERMOHYD has been developed to perform the steady state thermal-hydraulic calculations of the MTR type research reactors fueled with plate-type-fuel, in the ranges of pressures and temperatures typical for MTR reactors. The code calculates the coolant and clad surface temperatures along the cooling channel, as well as the critical heat flux and provides values for the critical phenomena parameters (RDR, ONBR and DNBR ratios) for reactor cores. The code takes into account the variation of the physical properties of the coolant with temperature. It also accounts for the uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations which have been considered in the analysis. The THERMOHYD code was tested against the well known PARET code and proved to be accurate and more conservative in some aspects.
Adaptation of the FEMAXI-6 code and RBMK fuel rods model testing employing the best estimate approach
Authors: Marao A. , Kaliatka T. , Kaliatka A. , Ušpuras E.
Published: December 2013

This paper provides information about the possibility to apply the FEMAXI-6 code for RBMK-1500 reactors. New thermal properties and correlations responsible for thermal analysis (thermal conductivity and heat capacity) according to RBMK-1500 specification were included into the FEMAXI-6 code. The RBMK-1500 fuel rod model using the adapted FEMAXI-6 code was developed and tested by employing uncertainty and sensitivity analysis. The processes in fuel rods during normal plant operation were modelled. The results received were compared with calculations performed by specialists from the Kurchatov Institute (designers of RBMK). The reasonable agreement of both calculation results shows that the adapted FEMAXI-6 code and the developed model are suitable for future analysis of processes in fuel rods of RBMK-1500 reactors. Also this work demonstrates what sensitivity and uncertainty analysis can be used for the model development.
Interactions of radionuclides in solution with ions and with trace concentrations of solid phase
Authors: Kepák F.
Published: December 2013

Interactions of radionuclides in extremely low concentrations were determined by measurement of their apparent self-diffusion coefficients. Formation of trace amounts of hydrolysis products, hydroxides of RuIV, MnIV, FeIII, EuIII and of ionic precipitate of AgI were determined from apparent self-diffusion coefficients. Transition of 85Sr2+ by ion exchange or by coprecipitation into trace amounts of FeIII-hydroxide and MnIV-hydrated oxide was revealed by this method. 85Sr2+ sorbed on trace amounts of FeIII-hydroxide and MnIV-hydrated oxides lost its individual chemical properties. This can negatively affect its separation from solution. Formation of unstable embryos of Ag131I trace precipitate in unsaturated solution at concentration of Ag+ 1 · 10−9 mol · l−1 and 131I−1 · 10−8 mol · l−1 was detected by self-diffusion. Formation of trace amounts of Ag131I led to the decrease of sorption of 131I− on hydrated ferric oxide.
Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor
Authors: Tusheva P. , Schäfer F. , Reinke N. , Altstadt E. , Rohde U. , Weiss F. P. , Hurtado A.
Published: December 2013

As a consequence of a total loss of AC power supply (station blackout) at a VVER-1000 leading to unavailability of major active safety systems, the safety criteria ensuring the safe operation of the nuclear power plant would be violated and core heat-up with possible core degradation could occur. A dedicated accident management measure (primary side depressurization) can be applied to reduce the primary pressure and to activate the injection from the passive emergency core cooling systems (accumulators). The analyses presented in this paper are aiming at both a detailed investigation of the accident sequence, taking into account the depressurization of the primary circuit, and the possibilities to prevent or at least to mitigate a damage of the reactor core so as to gain additional time for taking necessary countermeasures. The analyses are performed using the codes ASTEC and ATHLET developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH).
Numerical investigation of heat transfer in the vertical annulus between pressure tube and calandria tube of the advanced water cooled reactor
Authors: Vaidya A. M. , Borgohain A. , Maheshwari N. K. , Govindan D. , Vijayan P. K.
Published: December 2013

In advanced water cooled reactors, an annular gap exists between pressure tube and calandria tube. The gap is closed from top but is open from bottom. Due to differential temperature between pressure tube and calandria tube, air flow is induced by natural convection. This leads to heat transfer from pressure tube to calandria tube. The quantification of the heat transfer between pressure tube and calandria tube is numerically carried out with the help of the CFD code PHOENICS. Validation of the CFD code with experimental results and some established computational work from the literature has been done in order to verify the accuracy of the code. The natural convection phenomenon in the annular gap is then simulated. The velocity and temperature fields obtained from the CFD simulation are used to compute local and average heat transfer coefficients. Heat transfer coefficients for various pressure tube temperatures are computed. The effect of water on the heat transfer in the annular gas is also studied.
Definition of conservative conditions for RIA analysis in the modernized RBMK reactor core
Authors: Pabarcius R. , Tonkunas A.
Published: December 2013

During the past decade a number of modifications were implemented in the Ignalina NPP reactors cores. A significant reactor core loading change allowed increasing the safety level of the plant. During previously performed safety analyses a question has arisen concerning conservatism of used initial boundary conditions for investigation. The impact of different axial fuel burn-up and power-density profiles in the reactor core on modelling results as well as estimation of conservative conditions for the analysis of reactivity initiated accidents in the modernized RBMK-1500 reactor core were analyzed. The results of the investigations are presented in this paper. Simulation results show that previously used initial boundary conditions for modelling of accidents are changing due to a change of composition and loading in the modernized RBMK reactor core.
Research on pitting corrosion of steam generator heat transfer tubes based on acoustic emission
Authors: Hou S. , Luo J. , Zhang Q. , Xu J.
Published: December 2013

Corrosion of steam generator heat transfer tubes (SGHTT) is one of the important problems which affect safety operation of nuclear power plants (NPP), and the hazard of pitting corrosion of heat transfer tubes is the most serious. With an acoustic emission device, the signals during a corrosion test on SGHTT were collected and analyzed, and the corrosion points in the tubes were located accurately. The results show that pitting corrosion of heat transfer tubes has passed through three periods in its development: expansion phase, stationary phase and rapid developing phase. The corrosion damage of HTT can be found earlier with acoustic emission than any other non-destructive testing methods. Acoustic emission can be used for on-line and real-time monitoring of the safety and operation of the steam generator and has therefore a great significance.
UN approximation to critical slab problem for one-speed neutrons with isotropic, forward and backward scattering
Authors: Öztürk H.
Published: December 2013

A modified version of the UN method is used to solve the critical slab problem in the case of a combination of forward and backward scattering with isotropic scattering in a uniform finite slab. The Marshak boundary condition is used for the calculation of the critical slab thicknesses and it is shown that the method gives accurate results in one-dimensional geometry and converges rapidly with easily executable equations. The presented numerical results are compared with the results available in literature.
Integrated online condition monitoring system for nuclear power plants
Authors: Hashemian H. M.
Published: December 2013

Online condition monitoring or online monitoring (OLM) uses data acquired while a nuclear power is operating to continuously assess the health of the plant's sensors, processes, and equipment; to measure the dynamic performance of the plant's process instrumentation; to verify in-situ the calibration of the process instrumentation; to detect blockages, voids, and leaks in the pressure sensing lines; to identify core flow anomalies; to extend the life of neutron detectors and other sensors; and to measure the vibration of reactor internals. Both the steady-state or DC output of plant sensors and their AC signal or noise output can be used to assess sensor health, depending on whether the application is monitoring a rapidly changing (e.g., core barrel motion) or a slowly changing (e.g., sensor calibration) process. The author has designed, developed, installed, and tested OLM systems (comprised of software/hardware-based data acquisition and data processing modules) that integrate low-frequency (1 mHz to 1 Hz) techniques such as RTD cross-calibration, pressure transmitter calibration monitoring, and equipment condition monitoring and high-frequency (1 Hz to 1 kHz) techniques such as the noise analysis technique. The author has demonstrated the noise analysis technique's effectiveness for measuring the dynamic response-time of pressure transmitters and their sensing lines; for performing predictive maintenance on reactor internals; for detecting core flow anomalies; and for extending neutron detector life. Integrated online condition monitoring systems can combine AC and DC data acquisition and signal processing techniques, can use data from existing process sensors during all modes of plant operation, including startup, normal operation, and shutdown; can be retrofitted into existing PWRs, BWRs, and other reactor types and will be integrated into next-generation plants.
Nuclear data for cyclotron production of 114mIn/114In and 140Nd/140Pr used in gamma camera monitoring, RIT, ERT and PET
Authors: Sadeghi M. , Ghanbarzadeh A. , Enferadi M.
Published: December 2013

Aim of the presented study is to compare the calculated cross sections for the production of 140Nd/140Pr and 114mIn/114In via different reactions with incident particle energy up to 50 MeV as a part of systematic studies on particle-induced activations on metal target, theoretical calculation of production yield, calculation of target thickness requirement and suggestion for optimum reaction to produce Neodymium-140 and Indium-114m. 140Nd/140Pr and 114mIn/114In excitation functions via natCe(3He, xn)140Nd → 140Pr, 141Pr(p,2n)140Nd → 140Pr, 141Pr(d,3n)140Nd → 140Pr, 114Cd(p, n)114mIn → 114In, 114Cd(d,2n)114mIn → 114In and 116Cd(p,3n)114mIn → 114In reactions were calculated by ALICE/91, ALICE/ASH (GDH Model&Hybrid Model) and TALYS-1.2 (Equilibrium&Pre-Equilibrium) codes and compared to existing data.
Investigation of cross sections of reactions used in neutron activation analysis
Authors: Tel E. , Şahan M. , Uğur F. A. , Şahan H. , Aydin A.
Published: December 2013

In this study, neutron incident reaction cross sections for some target nuclei such as 24Mg, 27Al, 28Si, 56Fe, and 63Cu used in neutron activation analysis have been investigated. The new calculations on the excitation functions of 24Mg(n, p)24Na, 27Al(n, p)27Mg, 27Al(n, α)24Na, 28Si(n, p)28Al, 56Fe(n, p)56Mn, and 63Cu(n, 2n)62Cu reactions have been carried out for incident neutron energies up to 20 MeV. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the new geometry dependent hybrid model and the full exciton model. Equilibrium effects are calculated according to the Weisskopf–Ewing model. In the present work, reaction cross-sections have been calculated by using empirical formulas developed for energies of 14–15 MeV. The calculated results are discussed and compared with the experimental data taken from the EXFOR database.
CFD analysis of passive autocatalytic recombiner interaction with atmosphere
Authors: Gera B. , Sharma Pavan K. , Singh R. K. , Vaze K. K.
Published: December 2013

In water cooled power reactors, significant quantities of hydrogen could be produced following a severe accident (loss-of-coolant-accident along with non availability of Emergency Core Cooling System) from the reaction between steam and zirconium at high fuel clad temperature. In order to prevent the containment and other safety relevant components from incurring serious damage caused by a detonation of the hydrogen/air-mixture generated during a severe accident in water cooled power reactors, passive autocatalytic recombiners (PAR) are used for hydrogen removal in an increasing number of French, German and Russian plants. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces generating steam and heat. Numerous tests and simulations have been conducted in the past to investigate passive autocatalytic recombiners behaviour in situations representative of severe accidents. Numerical models were developed from the experimental data for codes like COCOSYS or ASTEC in order to optimise the passive autocatalytic recombiners location and to assess the efficiency of passive autocatalytic recombiners implementation in different scenarios. However, these models are usually simple (black-box type) and based on the manufacturer's correlation to calculate the hydrogen depletion rate. Recently, uses of enhanced CFD models have shown significant improvements towards modeling such phenomenon in complex geometry. The work presents CFD analysis of interaction of a representative nuclear power plant containment atmosphere with passive autocatalytic recombiners simulated using the commercial Computational Fluid Dynamics code for PAR Interaction Studies (PARIS benchmarks) exercise. A two-dimensional geometrical model of the simulation domain was used. The containment was represented by an adiabatic rectangular box with two PAR located at intermediate elevations near opposite walls. The flow in the simulation domain was modelled as single-phase. The results of the simulations are presented and analysed.
CFD simulation of thermal discharge behaviour in the Kadra reservoir at the Kaiga atomic power station
Authors: Sharma P. K. , Goyal P. , Markandeya S. G. , Ghosh A. K.
Published: December 2013

The thermal pollution arising out of discharge of hot water from the power plant condensers into the natural water bodies such as rivers, lakes, reservoirs, oceans etc. has been a serious concern to environmentalists ever since the plants started operating world over. In the past forty to fifty years, the methods of calculations for predicting the velocity and temperature fields in the affected regions of the stagnant/flowing water bodies have undergone a significant improvement. Currently, use of Computational Fluid Dynamics (CFD) codes for performing these calculations is gaining popularity. However, several factors such as the assumed computational domain and its discretisation, the boundary conditions used, representation of hydrodynamic characteristics (laminar/turbulent, buoyant/non-buoyant), etc. have a strong influence on the accuracy of predictions by such a model. A CFD code STAR-CD has been used for analyzing the thermal plume behaviour in the Kadra reservoir at Kaiga Atomic Power Station (KAPS). The predictions from these calculations of two units in operation have been found to be in good agreement with the site data made available from earlier studies. The present paper briefly describes the model developed using STAR-CD and results obtained for the Kadra reservoir at KAPS.
Modified UN method for the reflected critical slab problem with forward and backward scattering
Authors: Öztürk H.
Published: December 2013

The critical slab problem is investigated in one-speed neutron transport theory for reflecting boundary conditions using second kind Chebyshev polynomials approximation. The forward-backward-isotropic scattering kernel is used in a uniform homogeneous slab. The critical slab thicknesses for various values of the collision parameter, forward and backward scattering and reflection coefficient are given in the tables together with the ones available in literature and they are in good agreement.
Novel method to produce 109Cd via proton irradiation of electroplated silver on a gold-coated copper backing
Authors: Gholamzadeh Z. , Sadeghi M. , Mirzaei M. , Aref M.
Published: December 2013

Silver electrodeposition on gold layer was carried out by the alkaline plating baths to produce cadmium-109 via natAg(p, n)109Cd nuclear reaction. Gold was electrodeposited on copper backing in the beginning; the bath content consisted of 17.7 gl−1 KCN, 6.6 gl−1 Au, 6.6 gl−1 K2CO3 and 3.3 gl−1 Na2CO3 while acidity and temperature of the bath adjusted at 10 and 45°C respectively. A DC current density of ca 2.08 mA · cm−2 was used to electroplate gold. A gold layer of 63 μm thickness with suitable morphology was obtained after the electrodeposition. Silver was electrodeposited on the gold layer with 100% efficiency using a cyanide bath. The silver target was irradiated with 15 MeV proton beam and current of 150 μA; the 109Cd production yield was 2.0 μCi/μA · h (0.074 MBq/μA · h). The target material was dissolved by 14 M HNO3. Cadmium-109 in the dissolved target solution was separated from silver by evaporation process. Cadmium recovered with more than 88% efficiency involved non-detected silver impurity.
Simulation of load following mode of operation for a natural circulation pressure tube type BWR
Authors: Kumar R. , Gaikwad A. J. , Contractor A. D. , Srivastava A. , Lele H. G. , Vaze K. K.
Published: December 2013

A new nuclear power reactor under design study is a vertical pressure tube type boiling light water cooled and heavy water moderated. One of the passive design features of this reactor is the heat removal through natural circulation of primary coolant at all power level with no primary coolant pumps. Nuclear plants are mainly base load units, but the proposed plant with various advance features has to operate in load following mode i.e. Reactor follows Turbine (in a limited range). In this mode, any alteration in turbine load results in the steam pressure change. The steam pressure error is fed to the Reactor Regulating System (RRS), which changes the reactor power to control the system pressure. To study this mode of plant operation, a plant simulation model with the feedbacks from various controllers has been developed using the RELAP5 code. This integrated plant model has been used for simulating the load-varying scenario for a change in plant load. All the process dynamics, modeling, design verification and performance issues are discussed in this paper.
Atmospheric dispersion of accidental release of radioactive gases from high enriched and low enriched fuel of Miniature Neutron Source Reactors (MNSR)
Authors: Mahmood T. , Iqbal M.
Published: December 2013

Analysis has been done for accidental release of radioactive gases from high and low enriched fuel of miniature neutron source reactors (MNSR). Calculations have been performed for existing high enriched uranium (HEU) fuelled core and prospected low enriched uranium (LEU) fuelled core. Composition of existing HEU fuel is UAl4-Al while prospected LEU fuel is UO2. The ORIGEN2 code has been used for burn up and source term calculations. The “Hotspot” version 2.06 has been employed for radiological dispersion analysis to model the release of radioactivity in case of accident. Site specific meteorological data has been employed to determine the dose values for different stability classes as a function of downwind distance. In case of a radiological accident releasing radioactive gases at ground level, the maximum Total Effective Dose Equivalent (TEDE) is about 0.26 Sv for the most prevailing meteorological conditions at the site. There were no appreciable differences found in doses for HEU and LEU fuel.
Investigations of bi-directional flow behaviour in presence of a large vertical opening in a fire compartment
Authors: Sharma P. K. , Gera B.
Published: December 2013

In the complex thermal hydraulics codes developed for fire, reactor and containment safety the junctions in the multi-compartment geometries are often modeled as uni-directional junctions and some construct of flow coefficient. However, certain large size junctions are known to depict bi-directional flow behaviour under specific circumstances. The CFD based computer code FDS was used for an earlier reported study of fire in an enclosure on the bidirectional flow behaviour in present of a wall opening. Numerical simulation is directed to monitor the entrainment of the fresh air from outside to the fire compartment and resulting plume deflection due to presence of a big opening. The paper presents the details of the analysis, the results obtained, and comparison with the reported experimental data in terms of plume deflection, entrainment. Detailed investigations have been carried out to understand the bi-directionality of a junction by analyzing studying the outgoing hot air flow and incoming cold air.
Influence of corrosion processes on the head loss across ECCS sump strainers
Authors: Kryk H. , Hoffmann W. , Waas U.
Published: December 2013

Corrosion investigations were carried out using hot-dip galvanized steel samples in order to identify the corrosion mechanisms as well as the nature of corrosion products formed. Corrosion of grating treads may exacerbate ECCS strainer clogging problems during LOCA events in LWR. For the related experiments, tailored bench-scale facilities have been constructed. The results of batch experiments as well as of long-time corrosion investigations in the Korr VA test facility suggest that there is a multi-stage corrosion process. The first stage comprises the dissolution of the Zn layer in the coolant forming Zn2+ ions, which can generate crystalline zinc borate in boric acid solutions, if the ion concentration exceeds the saturation concentration. The base material (steel) is cathodically protected as long as the zinc layer is present. During the second stage, the base material dissolves forming insoluble corrosion particles, which can subsequently lead to acceleration of the strainer clogging. The main influences on corrosion were identified as impact of the water onto the corroding surface, water chemistry and Zn surface/coolant volume ratio.
Evidence of having control over loss-of-coolant accidents with release of insulation material under consideration of actions taken in German BWR
Authors: Schaffrath A. , Henschel M.
Published: December 2013

The following contribution describes the main features of the demonstration of the control of loss of coolant accidents with release of insulation material under consideration of actions taken in German BWR. The general criteria for the assessment as well as comparable individual aspects are adopted from the RSK statement (374th meeting) for PWR from July 2004. After the Barsebäck event in 1992 administrative regulations were implemented for securing the plant cleanliness as well as modifications to strainers of the ECCS and the containment (e.g. the introduction of perforated plates in the control rod chamber). This contribution illustrates that sufficient examinations and tests have been performed for German BWR for a final positive evaluation that the amount of fibers which enter the core remain below the amount at which core cooling is no longer guaranteed. Currently, for one BWR an additional optimation of the ECCS systems (here the introduction of a so-called “sandwich” wire cloths in the ECCS strainers) has currently been applied for. Through this the core entry of insulation material – even for fibers with a covering small lengths spectrum – can be further reduced and currently existing safety margins further increased.
Experiments on the integral test facility “Erlanger Wanne”
Authors: Ganzmann I. , Waas U.
Published: December 2013

The facility is designed to investigate processes following a loss of coolant accident (LOCA) in the region of the reactor sump as well as downstream the sump strainers. The following parameters have been considered: The debris transport and sedimentation behavior in the reactor sump region; pressure loss caused by debris agglomeration on the sump strainer; the influence of strainer geometry and size of the strainer openings on the pressure loss; back-flushing ability of sump strainers; pressure loss caused by debris bypassing the sump strainer (downstream effect); the influence of erosion and corrosion processes on the pressure loss behavior.

The experimental studies led to the development of an accepted procedure to handle a LOCA for KWU PWR plants. For the EPR™ the efficiency of the debris retention concept in case of a LOCA was demonstrated.
Development of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR
Authors: Rohde U. , Baier S. , Duerigen S. , Fridman E. , Kliem S. , Merk B.
Published: December 2013

The Light Water Reactor (LWR) dynamics code DYN3D is extended and adopted for the application to block-type High Temperature Gas-Cooled Reactor (HTGR). A procedure for the cross section generation for the HTGR core calculations was developed. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. A full core analysis of the reference simplified HTGR core is performed with DYN3D using macroscopic nodal cross sections provided by HELIOS.

The SP3 transport approximation is integrated into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method was developed for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes.

A 3D heat conduction module coupled with a channel-type coolant flow model is implemented into the code. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.
A complementary Doppler Broadening formalism and its impact on nuclear reactor simulation
Authors: Dagan R. , Becker B. , Danon Y.
Published: December 2013

The Boltzmann Transport equation is the governing formalism upon which simulations of nuclear reactors are performed, in particular when strong absorption or anisotropic scattering are significant. On the left (loss) hand side of the balance equation one finds the absorption and the scattering cross section Σa(E′), Σs(E′) respectively. Those cross sections are energy and temperature dependent i.e. Doppler Broadened. The scattering cross section appears explicitly again on the right (production) hand side of the equation in its differential form ∫∫0
∞Σ(E → E′); Ω → Ω′)dE dΩ. However, this term is commonly evaluated at 0 K and it does not account for the existing resonances which are the underlying characteristic for Doppler Broadening. Evidently one gets an inherent inconsistency between the integral and differential scattering cross section within the transport solver codes. In this study this missing Doppler Broadened formalism for the differential scattering cross section is introduced in its stochastic and deterministic form. The impact on core criticality is shown to be up to 600 pcm and the change in the nuclides' inventory significant, in particular the 239PU content can be changed by several percents.
Multiple condensation induced water hammer events, experiments and theoretical investigations
Authors: Barna I. F. , Ezsöl Gy.
Published: December 2013

We investigate steam condensation induced water hammer (CIWH) phenomena and present experimental and theoretical results. Some of the experiments were performed in the PMK-2 facility, which is a full-pressure thermalhydraulic model of the nuclear power plant of VVER-440/312 type and located in the Atomic Energy Research Institute Budapest, Hungary. Other experiments were done in the ROSA facility in Japan. On the theoretical side CIWH is studied and analyzed with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to RELAP5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. New features are the existence of multiple, independent CIWH pressure peaks both in experiments and in simulations. Experimentally measured and theoretically calculated CIWH pressure peaks are in qualitative agreement. However, the computational results are very sensitive against flow velocity.
Strategy, main stages and progress of the Ignalina Nuclear Power Plant decommissioning
Authors: Poskas P. , Simonis V. , Smaizys A. , Sirvydas A. , Ragaisis V. , Kilda R. , Zujus R.
Published: December 2013

Unit 1 and Unit 2 reactors of the Ignalina Nuclear Power Plant were shutdown in December 31, 2004 and December 31, 2009, respectively. Preliminary and Final Decommissioning Plans for the Ignalina Nuclear Power Plant Units 1 and 2 were prepared, the latter of which was approved by Ministry of Economy in 2005. The main points of Final Decommissioning Plan and the main Decommissioning Projects for the Ignalina Nuclear Power Plant are summarized in this paper. Particular attention is drawn to the spent nuclear fuel and radioactive waste management projects. Preliminary analysis showed that the main mass of equipment to be dismantled is concentrated in reactor buildings (∼39%) and turbine halls (∼37%) and consists from metals of different types (∼83%). From radiological point of view, a major part (∼61%) of contaminated equipment can be classified as very low level radioactive waste.
Environmental safety aspects of the new solid radioactive waste management and storage facility at the Ignalina Nuclear Power Plant
Authors: Ragaisis V. , Poskas P. , Simonis V. , Adomaitis J. E.
Published: December 2013

New solid radioactive waste management and interim storage facilities will be constructed for the Ignalina Nuclear Power Plant to support ongoing decommissioning activities, including removal and treatment of operational waste from the existing storage buildings. The paper presents approach and methods that have been used to assess radiological impacts to the general public potentially arising under normal operation and accident conditions and to demonstrate compliance with regulations in force. The assessment of impacts from normal operation includes evaluation of exposure arising from release of airborne radioactive material and from facilities and packages containing radioactive material. In addition, radiological impacts from other nearby operating and planned nuclear facilities are taken into consideration. The assessment of impacts under accident conditions includes evaluation of exposure arising from the selected design and beyond design basis accidents.
Externally organized exercises for nuclear emergency response – knowlegde gained from twenty years of practical work as a technical expert adviser for radiation protection
Authors: Löffler C. , Schumacher P.
Published: December 2013

At the end of the eighties the TÜV NORD established an expert consultation service for radiation protection in case of nuclear accidents. Since that time, educated and regularly trained surveyors also take part as expert advisors in out of plant emergency exercises for radiation protection in the federal countries Schleswig-Holstein, Hamburg and Mecklenburg-Western Pomerania. Their responsibility is to assist and to advise the emergency response commander regarding both the predicted and the current radiological situation. Furthermore, the technical expert adviser for radiation protection communicates with other expert advisers (agent of the operator, medical staff, and meteorologists) as well as with measuring and emergency services. The experience gained during emergency exercises is summarized and documented. In the course of the past twenty years the implementation of central structures, the software systems for determination of the radiological impact and the change of communication systems have especially proved to be successful. For the future it can be assumed that exercises covering several days will be performed more often, so that not only the accident process is simulated correctly but also the conditions for the personnel of each disaster prevention unit can be simulated more realistically.
Principles of off-site nuclear emergency exercises
Authors: Miska H.
Published: December 2013

Due to high safety standards at nuclear power plants, no experience exits with nuclear emergencies in Western Europe. Thus, emergency exercises are the only possibility to assure effective protective measures should the very unlikely severe accident occur. The main objectives of exercises are generally the check of response plans for suitability, the test of the equipment's applicability and training of personnel for the unusual task to manage a nuclear emergency. After an introduction into the different types of exercises, this contribution focuses on off-site nuclear emergency exercises, explaining frame conditions to ensure good practice and, finally, reports some experience from exercises.
Switzerland: a culture of preparedness
Authors: Storch D. , Kenzelmann M.
Published: December 2013

The National NBC Protection and Coordination Office supports the activities of all members of the National NBC protection network. These include the development and implementation of operational principles, as well as training and pro-tection concepts on behalf of the head of the Steering Committee on Radioactivity (LAR) and the Federal Office for Civil Protection (FOCP), the chair of the Federal Commission for NBC Protection (ComNBC) and the chair of the Cantonal NBC Coordination Platform (KPABC). Besides exercises at the political-strategic level (civilian and military), operational training exercises are held on a regular basis.
Systematic preparation, execution and evaluation of emergency exercises at the Beznau nuclear power plant
Authors: Tenschert J.
Published: December 2013

Based on federal acts and a specific guideline of the nuclear authority ENSI, strict requirements are imposed on emergency exercises at Swiss NPPs. The Beznau NPP has conducted emergency exercises for more than 30 years. Systematic exercise planning assures that all emergency cases defined in the plant-specific emergency plan are considered in the exercise scenarios. Technically oriented scenarios cover all groups of initiating events and all safety levels of the defense in depth principle. The exercise results are an important input for optimization measures in the areas of emergency organization, documentation and infrastructure. Due to the goal-oriented enforcement of laws and guidelines by the nuclear authority ENSI, emergency exercises serve as a motor of further optimization of emergency preparedness.
Development of program DETSIM to simulate detector's full energy peak efficiency
Authors: Thanh T. T. , Tao C. V. , Au B. H. , Chuong H. D.
Published: December 2013

In this paper, the new software named DETector SIMulation (DETSIM) was developed from PENELOPE code system method to simulate gamma spectra and calculating full energy peak efficiency (FEPE). The new software is using graphic user interfaces for easy-to-use purpose. As a first application post test calculations has been performed for an experiment of a mixed source (241Am, 109Cd, 57Co, 139Ce, 51Cr, 113Sn, 85Sr, 137Cs, 60Co and 88Y) located at two positions away from detector window and energy ranges between 60 keV and 2000 keV. The results showed a general agreement between the simulated efficiencies and the experimental data. The simulation results are typically 7% higher than the experimental data in the low energy range. This means that our simulation program is good enough for later studies on our HPGe spectrometer.
Exercises of first responder organisations in Austria
Authors: Schönhacker S.
Published: December 2013

Although there is no nuclear industry in Austria, there are still many possible scenarios which require proper preparation through exercises. The legislative basis for the interventions of the police are the radiation protection law and the upon based interventions regulation, furthermore the penal law and the law on the transport of dangerous goods. The fire brigade has federal fire fighting laws and internal regulations as a regulatory basis. Exercises of first responder organisations take place once a year at least; the scenarios reflect the actions intended by the regulations. Aeroradiometry is a special technique conducted by the police, while the fire brigade may bring heavy equipment to use. Further improvement of the cooperation of different first responder organisations is a major goal of combined exercises.
Best estimate plus uncertainty analysis of LBLOCA for Indian PHWR
Authors: Srivastava A. , Trivedi A. K. , Lele H. G. , Munshi P. , Vaze K. K.
Published: December 2013

Deterministic safety analysis is an important tool for confirming the adequacy of provisions within the defense-in-depth concept for the safety of nuclear power plants. One of the important design basis events is considered to be a complete double-ended guillotine rupture i.e. Loss of Coolant Accident (LOCA) of largest and coldest pipe in the primary coolant circuit. The present work deals with this scenario for an Indian PHWR. It highlights the identification of critical break size leading to the maximum clad temperature using the best estimate code RELAP5. Further, important parameters affecting the clad temperature are described along with results of initial sensitivity studies to select dominant uncertain parameters. For the uncertainty propagation, Latin Hypercube Sampling (LHS) is used instead of simple random sampling for Monte-Carlo simulation. The inherent characteristic of LHS is to reduce the required runs for Monte-Carlo simulation to manageable order for the current computing capability. The 95th percentile value of peak cladding temperature (PCT) is obtained by the method and compared with acceptance criteria.
Dosimetric aspects of 103Pd radioactive stent source
Authors: Sadeghi M. , Kiavar O.
Published: December 2013

This study aims to determine AAPM TG-60 dosimetric quantities in regions surrounding the 103Pd stent wall by MCNP5 Monte Carlo code. The Palmaz-Schatz stent was modeled by a hollow cylinder of 17.89mm length (2mm diameter) with net surface very similar to real stent. The Dose deposited per photon (Gy), relative dose, Anisotropy function, F(r,θ), and radial dose function, gL(r), were described at AAPM TG-60 protocol were generated from these values and listed in tabular format. For benchmarking, the relative dose values were verified with TG-43 and EGS4 code results at identical conditions, relative to the radial distances from surface of the stent. There were noticeable results. These physical dosimetric parameters can be used in future treatment planning systems for IVBT.
Radiological significance of coal, slag and fly ash samples from the Eastern Black Sea region
Authors: Damla N. , Cevik U. , Kara A.
Published: December 2013

This work presents a study of natural radioactivity levels in coal and its combustion residues (fly ash and slag) used in the houses in Black Sea Region, Turkey. Coal, fly ash and slag samples were provided from different locations of the region and analyzed by gamma spectroscopy using a high-purity germanium detector (HPGe). Also, chemical analyses of these samples were carried out using energy dispersive x-ray fluorescence spectrometer. The mean 226Ra activity concentrations in coal, slag and fly ash were measured as 83, 99 and 38Bqkg–1, respectively. The mean 232Th activity concentrations in coal, slag and fly ash were measured as 108, 113 and 50Bqkg–1, respectively. The mean 40K activity concentrations in coal, slag and fly ash were found to be 366, 381 and 204Bqkg–1, respectively. The potential radiological hazards associated to these materials were evaluated by calculating the radium equivalent activity (Raeq), the air absorbed gamma dose rate (D), the annual effective dose rate (AED), the external hazard index (Hex) and internal hazard index (Hin) and compared with the internationally accepted or reference values. The mean Raeq values of the coal, fly ash and slag samples were lower than the recommended maximum values 370Bqkg–1 by the Organization for Economic Cooperation and Development (OECD). The overall mean outdoor terrestrial gamma air absorbed dose rate in coal, fly ash and slag samples are 119, 129 and 62nGyh–1 and the corresponding outdoor annual effective doses are 0.60, 0.32 and 0.64mSvy–1, which is higher than the worldwide average (0.07mSvy–1), respectively. Moreover, the enrichment factors relative to the input coal are calculated for the radionuclide contents observed. Calculated enrichment factor values for 226Ra and 232Th were found 1.14 and 1.01, respectively.
Computational models to dertermine fluiddynamical transients due to condensation induced water hammer (CIWH)
Authors: Swidersky H. , Schaffrath A. , Dudlik A.
Published: December 2013

Condensation induced water hammer (“condensation hammer”, CIWH) represent a dangerous phenomenon in pipings, which can endanger the pipe integrity. If they cannot be excluded, they have to be taken into account for the integrity proof of components and pipe structures. Up to now, there exists no substantiated model, which sufficiently determines loads due to CIWH. Within the framework of the research alliance CIWA, a tool for estimating the potential and the amount of pressure loads will be developed based on theoretical work and supported by experimental results. This first study discusses used computational models, results of experimental observations and gives an outlook onto future techniques.
Delayed equilibrium model and validation experiments for two-phase choked flows relevant to LOCA
Authors: Bartosiewicz Y. , Seynhaeve J.-M. , Serre G.
Published: December 2013

This paper deals with the 1-D Delayed Equilibrium Model (DEM) for choked or critical flow rate in steady state or quasi-steady state conditions and the selection of the relevant experimental data for assessing such models. In particular, the focus is made on thermodynamic non-equilibrium conditions, which prevail in the flashing process near the critical section. In this regard, relaxation models such as the DEM developed and tested in previous studies at UCL was revisited and improved in view of their implementation in the next CATHARE code generation during the EU NURISP (NUclear Reactor Integrated Software Platform) project. A methodology to implement the DEM into is developed. Some new results of the DEM are compared against experimental data such as Super Moby-Dick experiments done in CEA and the well-known Marviken experiments performed at quasi-real scale geometry of Nuclear Plant.