bluuebunny/crossref_metadata_2025_split huggingface dataset:
Definition of conservative conditions for RIA analysis in the modernized RBMK reactor core
Authors: Pabarcius R. , Tonkunas A. 
Published: December 2013

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.