bluuebunny/crossref_metadata_2025 huggingface dataset:
A discrete ordinates scheme for void fraction evaluation with nonstandard reflective conditions and weakly divergent beams
Authors: de Abreu M. P.
Published: December 2013

In this article, a number of due changes in both a discrete ordinates neutron transport model and a companion void fraction evaluation scheme are introduced. These changes are due to the explicit consideration of a weakly divergent neutron beam, i. e. a neutron beam consisting of a monodirectional component with normal incidence, which is modeled by a Dirac delta distribution, and an angularly continuous component, which is modeled by a smooth function of the angular variable. Computational tests are performed in order to illustrate the numerical consistency of the evaluation scheme for weakly divergent beams with respect to the order of angular quadrature, as well as its low sensitivity to experimental inaccuracies in the detector responses.
Methodology for Fire PSA during design process
Authors: Blombach J. , Kollasko H.
Published: December 2013

Fire PSA is an essential part of a full scope level 1 PSA. Cable fires play an important role in fire PSA. Usually, cable routing is therefore modeled in detail. During the design of new nuclear power plants the information on cable routing is not yet available. However, for the use of probabilistic safety insights during the design and for licensing purposes a fire PSA may be requested. Therefore a methodology has been developed which makes use of the strictly divisional separation of redundancies in the design of modern nuclear power plants: cable routing is not needed within one division but replaced by the conservative assumption that all equipment fails due to a fire in the concerned division; critical fire areas are defined where components belonging to different divisions may be affected by a fire. For the determination of fire frequencies a component based approach is proposed. The resulting core damage frequencies due to fire are conservative.
The Open PSA Initiative for next generation probabilistic safety assessment
Authors: Epstein S. , Rauzy A. , Reinhart F. M.
Published: December 2013

The Open PSA Initiative round table discussion session is to bring the PSA community up-to-date on our meetings, workshops, and working groups during the last year. We will have members of the initiative speak on their point of view of the initiative, what they have gained, and where we as a group want to go during the next year. The main idea is to use appropriate modules of different computer codes for a specific purpose which requires a common platform and a common PSA format. In this short paper we introduce our main project for this year: the Open PSA Model Exchange Format.
Overview on the different applications of probabilistic safety assessment for nuclear power plants
Authors: Berg H.-P.
Published: December 2013

Worldwide it can be recognised that the use of probabilistic safety assessment (PSA) in regulatory as well as operational decision-making is state of the art and seen as a successful development. Therefore, in most cases the regulator encourages the performance of PSAs to provide information to complement and support the defence in depth philosophy as well as operational configuration decisions. The main application of the PSA is still as part of integrated safety reviews, in particular in the frame of comprehensive (periodic) safety reviews. Other more specific applications areas of PSA are, among others, design evaluation, event analysis with aid of PSA, evaluation of technical specifications; risk-informed in-service inspection, risk monitoring and accident management. The extent of these applications vary from country to country but has been increasing during the last years.
Practicle example of the defence of a contamination incident
Authors: Höfer H.
Published: December 2013

By routine examinations of a radiation worker, a contamination incident with transuranium nuclides was discovered. Immediate investigations showed a serious degree of the incident. First countermeasures were done to narrow the effects down and to avoid a spread of contamination. A complex decontamination followed for which the radiation protection infrastructure had to be built up.
Background information about polonium-210 – considerations on its biokinetics and internal dosimetry
Authors: Oeh U. , Li W. B. , Gerstmann U. , Giussani A. , Paretzke H. G.
Published: December 2013

The mysterious death of the former secret service agent Alexander Litvinenko on 23 November 2006 suddenly attracted the attention of the public to the rather unknown radionuclide 210Po. In this respect, this paper presents some general background information about this radionuclide as well as some details about its biokinetic behaviour and internal dosimetry. The results are discussed in the context of the Litvinenko case with regard to possible health effects on people from Germany being potentially contaminated. However, no evidence for contamination with 210Po could be found for any people tested at the GSF-National Research Center for Environment and Health (since 01.01.2008 Helmholtz Zentrum München).
Hazard and precautionary measures in connection with negligent or wilful contamination
Authors: Ahlfänger R. , Scheidegger R.
Published: December 2013

Handling open radioactive substances requires extensive precautionary measures in order to avoid any danger. During the past 100 years many persons were injured or killed due to unintended external irradiation, contamination of the skin or uptake of radioactive substances. With the help of organisational measures with users of sources and with operators of radionuclide laboratories or nuclear installations numerous accidents can be avoided or their consequences can be mitigated. Also the government itself is challenged in order to register and secure radioactive sources and maintaining the national safeguard. Education of personnel handling radioactive substances plays an important role when it comes to precautionary measures. In combination with suitable technical equipment it is possible to obtain a high degree of protection against the hazard from negligent or wilful contamination.
Prediction of time dependent standby failure rates for periodically tested components taking into account the operational history
Authors: Šimić Z. , Vrbanić I. , Vuković I.
Published: December 2013

The prediction of time dependent standby failure rates was studied, taking into account the operational history of a component. These studies are important for applications such as system modeling in probabilistic safety analysis to evaluate the impact of equipment aging and maintenance strategies on the risk measures (e. g. reactor core damage frequency) considered. The time dependent model for the standby failure rate is defined based on the Weibull distribution and the principles of proportional age reduction by equipment overhauls. The parameters which determine the standby failure rate are estimated, including the definition of the operational history model and likelihood function for Bayesian analysis of parameters for periodically tested standby repairable components. The operational history is provided as time axis with defined times of overhauls, surveillance tests and failures. Assessment of time dependent unavailability due to the failure of periodically tested standby components is described. As an example, the prediction of the future behavior of components for seven different operational histories is described.
Numerical simulation of passive catalytic recombiner
Authors: Gera B. , Sharma Pavan K. , Singh R. K. , Ghosh A. K.
Published: December 2013

Resolving hydrogen related safety issues, pertaining to nuclear reactor safety have been an important area of research world over for the past decade. Studies on hydrogen transport behavior and development of hydrogen mitigation systems are still being pursued actively in various research labs, including Bhabha Atomic Research Centre (BARC), in India. The Passive Catalytic Recombiner is one of such hydrogen mitigating devices consisting of catalyst surfaces arranged in an open-ended enclosure. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces and the heat of the reaction produces a natural convection flow through the enclosure. The present study aims for performing a CFD simulation to obtain the temperature distribution inside the recombiner. A 3D CFD model has been developed to study the mechanism of catalytic recombination and has been tested for literature quoted experiments. A parametric study has been performed for a particular recombiner geometry for various inlet conditions. Salient features of the simplified CFD model developed at BARC and results of the present model calculations are presented in this paper.
An analytical solution for the one-dimensional time-dependent S

N
transport equation for bounded and unbounded domains in cartesian geometry
Authors: Segatto C. F. , Vilhena M. T. , Gonçalez T. T.
Published: December 2013

In this work a general analytical solution is reported, expressed in integral form for the time-dependent, one-dimensional SN transport equation in cartesian geometry valid for bounded and unbounded domains (0 < x <), using the double Laplace transform technique. The main idea consists in the application of the Laplace transform technique in time variable and the solution of the resulting equation by the LTSN method using appropriate boundary conditions for bounded and unbounded domain problems. We also report about the numerical simulations carried out.
Design optimization of shell-and-tube heat exchangers using single objective and multiobjective particle swarm optimization
Authors: Elsays M. A. , Aly M. Naguib , Badawi A. A.
Published: December 2013

The Particle Swarm Optimization (PSO) algorithm is used to optimize the design of shell-and-tube heat exchangers and determine the optimal feasible solutions so as to eliminate trial-and-error during the design process. The design formulation takes into account the area and the total annual cost of heat exchangers as two objective functions together with operating as well as geometrical constraints. The Nonlinear Constrained Single Objective Particle Swarm Optimization (NCSOPSO) algorithm is used to minimize and find the optimal feasible solution for each of the nonlinear constrained objective functions alone, respectively. Then, a novel Nonlinear Constrained Multobjective Particle Swarm Optimization (NCMOPSO) algorithm is used to minimize and find the Pareto optimal solutions for both of the nonlinear constrained objective functions together. The experimental results show that the two algorithms are very efficient, fast and can find the accurate optimal feasible solutions of the shell and tube heat exchangers design optimization problem.
Neutron transmutation doping conceptual design
Authors: Osman M. M. , Agamy S. A. , Nagy M. S. , Sultan M.
Published: December 2013

This paper describes and investigates upgrading of the silicon doping facility at the ETRR-2 research reactor to accommodate larger size silicon ingots in the two positions available for irradiation in the thermal column of the reactor. Two methods of irradiation are proposed. Both are based on irradiating two ingots simultaneously in one position. After a certain irradiation time, the silicon ingots are reshuffled inside the irradiation channel. The results obtained show that the proposed irradiation strategies of doping silicon ingots are simple to handle and result in acceptable values of resistivity variation.
Chebyshev acceleration of criticality calculations in boundary element applications to neutron diffusion
Authors: Engin O. , Ozgener B.
Published: December 2013

In this study, Chebyshev polynomial method has been applied in the acceleration of the outer iterations for the numerical solution of the two dimensional neutron diffusion equation by the boundary element method. Boundary element discretization of the diffusion equation results in a full and nonsymmetric coefficient matrix which is quite different than the sparse and symmetrical matrices of the finite element and finite difference methods to which Chebyshev acceleration has been classically applied. To assess the merit of the Chebyshev method in the case of boundary element discretization of the multigroup diffusion equation constitutes the main objective of the study. Numerical experimentation has established that the Chebyshev acceleration is effective also in case of boundary element discretization despite the dissimilarity of the coefficient matrices compared to the more conventional methods.
Analysis of the instability of OTSG of movable nuclear power devices by using the multivariable frequency domain method
Authors: Hou S.-X. , Zhang Q.-H. , Luo J.-J. , Xu J. , Tian T.
Published: December 2013

A new method for analyzing the problem of the thermal-hydrodynamic instability of OTSG (Once-Through Steam Generator) of movable nuclear power devices is presented in this paper. The method is based on multivariable frequency domain theory and is more efficient for analyzing the instability of OTSG with coupling interactions and complicated boundary conditions than the usually used single variable method. A mathematical model for the system is derived from the fundamental equations by using perturbation, Laplace-transform and the nodalization techniques. The stable boundary and parametric effects on the stability of the system are evaluated with a computer code. Numerical examples are given and the predictions of the model are in good agreement with the experimental results.
In-reactor measurements of fuel centerline temperature variation during power change
Authors: Horhoianu G. , Ionescu D. V. , Pauna E. I.
Published: December 2013

The thermal response of the experimental fuel elements is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by simultaneous liberation of fission gases and pellet thermal expansion densification, swelling and relocation of fuel fragments. This paper describes briefly measuring techniques developed and currently used in INR Pitesti, and presents and discusses selected experimental results obtained during in-pile experiments performed in the TRIGA SS reactor of INR Pitesti in order to measure the fuel centerline temperature variation during power change.
Kritikalitätssicherheit der Endlagerung ausgedienter Kernbrennstoffe
Authors: Neuber J. C.
Published: December 2013

Kurzfassung

In dem nachfolgenden Beitrag werden die Grundsätze der Kritikalitätssicherheitsauslegung eines Endlagers ausgedienter Kernbrennstoffe erläutert. Diese Erläuterungen umfassen:

Sicherheitsprinzipien der Kritikalitätssicherheitsauslegung eines Endlagers

Konsistenz der Sicherheitsprinzipien beim Übergang von der Betriebsphase in die Nachbetriebsphase des Endlagers

Die prinzipielle Abhängigkeit der Länge erforderlicher Nachweiszeiträume von den eingelagerten Nuklidinventaren und deren zukünftigen Entwicklungen, die durch den radioaktiven Zerfall sowie durch mögliche langzeitliche Entwicklungen des Endlagers bedingte Einwirkungen bestimmt werden

Wahrscheinlichkeit des Eintretens von Kritikalität hinsichtlich bisher erwogener Einlagerungskonzepte und die Notwendigkeit des Nachweises der Sicherheit des Endlagers gegen potentielle Kritikalität in der Nachbetriebsphase

Die sich zur Beantwortung der Frage der Sicherheit des Endlagers gegen potentielle Kritikalität für die Kritikalitätsanalyse ergebenden Aufgaben werden skizziert. Die Bearbeitung dieser Aufgaben macht die Modellierung komplexer Mechanismen (Szenarien), die zu Kritikalität führen können, die Modellierung der komplexen und verschiedenartigen Kritikalitätsexkursionen und die Modellierung der Auswirkungen solcher Exkursionen auf die Sicherheit des Endlagers erforderlich. Die Entwicklung entsprechender Modelle erfordert die konzertierte Zusammenarbeit vieler Fachdisziplinen. Es wird angeregt, die bereits bestehenden Kooperationen unter Einbeziehung der Kritikalitätsanalyse in einer nationalen Initiative zum Projekt „Endlagerung“ noch näher miteinander zu vernetzen. Aufgabe einer solchen Kooperation wird es auch sein, die Risiken aller Tätigkeiten, Einrichtungen und Randbedingungen, die mit der Abfallkonditionierung sowie der Beladung, dem Betrieb und der Nachbetriebsphase eines Endlagers verbunden sind, zu bewerten und so optimal auszubilanzieren, dass das Gesamtrisiko dem Minimierungsgebot genügt. Dies schließt eine Bilanzierung des Risikos einer potentiellen Kritikalität in der Nachbetriebsphase gegen die Risiken ein, die mit der Durchführung von Maßnahmen verbunden sind, die zur physikalisch sicheren Verhinderung des Eintretens von Kritikalität in der Nachbetriebsphase vor und während der Einladung in das Endlages erforderlich wären.
Uncertainty analysis for fission products transport in CANDU primary heat transport during a severe accident
Authors: Apostol M. , Constantin M. , Leca A.
Published: December 2013

The work realized under the Severe Accident Research Network of excellence (SARNET) project has shown that the SOPHAEROS module, part of Accident Source Term Evaluation Code (ASTEC) can be fully used to simulate the fission products transport and deposition phenomena in the CANDU Primary Heat Transport (PHT) system. This paper presents an uncertainty analysis for the fission products transport in the CANDU PHT system during a severe accident to obtain the domains of the output parameters, for this study masses of Caesium, Strontium and Iodine deposited in the PHT system and its nodes, taking into account the associated input parameters uncertainties. Five uncertain parameters, the starting time for the releasing process, the duration of the releasing process, the releasing fractions for Cs, Sr and I have been chosen. To generate aleatory values for the uncertain parameters, a method and software have been developed and Monte Carlo simulations to determine uncertainties propagation through the SOPHAEROS module has been carried out.
Axial-power-shape induced frequency shifts in BWR stability measurements
Authors: Nylén H.
Published: December 2013

Higher reactor-core-resonance frequency than expected was recognised at the 1800 MWt BWR Barsebäck 2 in Sweden during stability measurements in 2004, and has subsequently been analysed in some detail. Estimation of the void velocity was made by analysing the phase shift between two levels in the LPRM-detector strings, and compared to calculations. Discrepancies in calculated frequency response between simulated perturbation and results from noise analysis were observed, whereas good agreement between calculated and measured void transport was found. The observed deviations in frequency calculations are most likely affected by the pronounced downward shift in power during the stability measurements, and in such circumstances, the shift in reactor-core-resonance frequency can be estimated from void transport calculations.
Numerical simulation on a HPLWR fuel assembly flow with one revolution of wrapped wire spacers
Authors: Kiss A. , Laurien E. , Aszódi A. , Zhu Y.
Published: December 2013

Three dimensional computational-fluid-dynamics simulations are performed for the fluid flow within a 40 rod fuel bundle in a square arrangement with a central moderator channel. To ensure spacing between the rods, the design of the bundle uses thin wires wrapped counter-clockwise around each rod. This geometry is presently investigated in the framework of the European High-Performance Light-Water Reactor (HPLWR), which operates at supercritical pressure of 25 MPa. A section with one revolution located in the evaporator region of the HPLWR core is investigated using hydraulic (to ensure fully developed flow inlet boundary conditions and reference for heated cases) and thermal-hydraulic boundary conditions. The geometry of wrapped wires gives rise to additional mixing and a circulating or ‘sweeping’ flow near the outer and inner regions of the fuel element next to the wall of the so called fuel assembly and moderator box. Some interesting flow features associated with the complex three-dimensional flow with significant transverse velocity components are visualized as the first evaluated result of this diversified investigation.
Neutronic and burn-up calculations of heterogeneous Thorium/Uranium fuel in pressurized water reactors
Authors: Foad B. , Mansour H. , Aziz M.
Published: December 2013

The purpose of this work is to study the feasibility of the Thorium/Uranium fuel cycle in heterogeneous Pressurized Water Reactors (PWR) core design. This paper focuses on the neutronic and burn-up analysis of the Thorium/Uranium fuel using the computer codes MCNPX and WIMS. The design is based on the Whole Assembly Seed and Blanket (WASB) concept, in which the individual seed (Uranium) and blanket (Thorium-Uranium) units occupy one full-size PWR assembly in a checkerboard core configuration. The results of the present models were compared with the solution of benchmark problems and satisfactory agreement was found.
Steam drum process dynamics and level control of a pressure tube BWR
Authors: Gaikwad A. J. , Vijayan P. K. , Bhartiya S. , Kumar R. , Lele H. G. , Sinha R. K.
Published: December 2013

The Advanced Heavy Water Reactor (AHWR) is a pressure tube type light water cooled heavy water moderated Boiling Water Reactor (BWR) with natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the Main Heat Transport (MHT) system configuration. These multiple (four) interconnected loops influence the Steam Drum (SD) level control adversely. Such a behavior has not been reported in the open literature. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Events (PIEs). This does not allow the partitioning of the Common Reactor Inlet Header (CRIH). If partitioning of the CRIH into four segments is allowed then, it will make each loop independent. Then the SD level control problems subside as the unaccounted interaction among the loops is eliminated. It has also been observed that the open loop response is stable, non-oscillatory and non-diverging for a step change in the feed flow rates. A conventional individual 3-element SD level controller cannot account for the highly coupled and interacting behaviors, of the four loops and SD levels. To overcome these interactions it is proposed to interconnect all the four steam drums in the liquid and vapor regions respectively. The influence of the interconnect configuration and the level controller are studied in detail to find a robust solution. The response obtained for unsymmetrical core power disturbances shows that the SD levels do not diverge and quickly settle to the set points assigned with SD interconnect. The proposed scheme also works well for most of the PIEs considered.
Rapid preparation of Uranium and Thorium alpha sources by electroplating technique
Authors: Hao Le Cong , Van Tao Chau , Van Dong Nguyen , Nghi Ho Nha , Dung Pham Thuy , Van Thong Luong
Published: December 2013

In this study, a simple procedure for the radiochemical determination of thorium isotopes and a robust technique for rapid preparation of uranium and thorium alpha source by electro-deposition using organic buffer solution is described. From the experiments, experience has shown that (1) TOPO is one of the best solvent for the extraction of thorium, (2) isopropyl alcohol is a suitable buffer solution for the electrodepositing of uranium; while acetone is the best choice in case of thorium, (3) the possibility of using of organic phase; containing thorium in the final chemical separation step for electrodepositing, (4) the use of only 3–5 ml of isopropyl alcohol or acetone; 3–5 mm of anode-cathode distance and relatively current 4–15 mA for electro-deposition procedure, (5) the actinides can be deposited quantitatively on a stainless steel disk within 20 to 40 min with nearly 100% of electro-deposition efficiency and the energy resolution of alpha-ray spectra of the electrode-posited sources obtained from alpha spectrometry are within the range of 37–50 KeV (FWHM).
Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling
Authors: El-Morshedy S. El-Din
Published: December 2013

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested.
Preliminary evaluation of a radioactive waste repository safety performance by a Monte Carlo simulation-based reliability model
Authors: Cadini F. , Avram D. , Zio E. , Luce A. , Taglioni A.
Published: December 2013

Disposal facilities are designed to provide long-term isolation of the wastes from the human environment by means of a system of barriers, both engineered and natural. In this regard, it is required that the containment function of the multi-barrier system be evaluated by a quantitative risk analysis procedure, also called performance assessment. This entails the implementation of a Probabilistic Safety Assessment model for representing the stochastic failure behavior of the repository barriers, estimating the release rates of radionuclides into the groundwater stream and the doses from subsequent ingestion of drinking water. To this aim, a Monte Carlo simulation scheme is provided within a reliability modelling approach for the safety performance analysis of a repository, accounting also for barrier degradation processes. An application to a case study of the literature is presented to validate the approach and demonstrate its simplicity and flexibility.
Inverse problems using Artificial Neural Networks in long range atmospheric dispersion
Authors: Sharma P. K. , Gera B. , Ghosh A. K.
Published: December 2013

Scalar dispersion in the atmosphere is an important area wherein different approaches are followed in development of good analytical models. The analyses based on Computational Fluid Dynamics (CFD) codes offer an opportunity of model development based on first principles of physics and hence such models have an edge over the existing models. Both forward and backward calculation methods are being developed for atmospheric dispersion around NPPs at BARC. Forward modeling methods, which describe the atmospheric transport from sources to receptors, use forward-running transport and dispersion models or computational fluid dynamics models which are run many times, and the resulting dispersion field is compared to observations from multiple sensors. Backward or inverse modeling methods use only one model run in the reverse direction from the receptors to estimate the upwind sources. Inverse modeling methods include adjoint and tangent linear models, Kalman filters, and variational data assimilation, and neural network. The present paper is aimed at developing a new approach where the identified specific signatures at receptor points form the basis for source estimation or inversions. This approach is expected to reduce the large transient data sets to reduced and meaningful data sets. In fact this reduces the inherently transient data set into a time independent mean data set. Forward computations were carried out with CFD code for various cases to generate a large set of data to train the Artificial Neural Network (ANN). Specific signature analysis was carried out to find the parameters of interest for ANN training like peak concentration, time to reach peak concentration and time to fall. The ANN was trained with data and source strength and locations were predicted from ANN. The inverse problem was performed using the ANN approach in long range atmospheric dispersion. An illustration of application of CFD code for atmospheric dispersion studies for a hypothetical case is also included in the paper.
Application of the TN
method to critical slab problem for one-speed neutrons with forward and backward scattering and efficiency of reflection coefficient
Authors: Bülbül A. , Kara A. , Anli F.
Published: December 2013

The one-speed neutron transport equation which includes isotropic forward and backward scattering has been studied using the TN approximation. Critical thicknesses for different values of the reflection coefficient R and c values have been calculated. The study shows that the TN method gives nearly the same results as the PN approximation in one-dimensional geometry. The results obtained in this study are presented in tables and are compared with PN approximation results.
LOCAs with release of insulation material: generic aspects and international solutions
Authors: Bröcker A. , Pointner W.
Published: December 2013

In case of a LOCA within the containment the released water/steam jet can fragment insulation material from the leaking pipe and from pipes in the neighborhood. The released insulation material and other materials like e.g. dust and paint chips are washed into the sump. A part of this debris reaches the sump strainers after starting ECCS pumps. Some material will be retained by the strainers depending on the mesh size or hole diameter of the strainers. The residual material penetrates the strainers and reaches downstream components and especially fuel elements. Therefore it has to be demonstrated that fuel elements are sufficiently cooled after LOCAs with release of insulation material and other materials. Several events since 1992 have shown the possibility of strainer clogging. This article discusses generic aspects and some international approaches to sump clogging issue and related tasks. An overview of the status quo and comparison of different approaches and background is given.
Multi-scale coupled code systems: from coarse-mesh to high-fidelity LWR core calculations
Authors: Périn Y. , Klein M. , Pasichnyk I. , Velkov K. , Seubert A. , Pautz A.
Published: December 2013

In order to perform safety oriented transient analyses, the entire nuclear power plant has to be simulated in most cases. However, a high level of resolution is only needed where the safety parameters may reach the safety criteria. In this paper, three scale resolutions are defined (NPP level, core level, fuel pin level). Thermal-hydraulic and neutron kinetic codes are already available for each scale. By coupling them, it is possible to develop a complete simulation tool giving access to the required level of precision at each scale level. For this purpose the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH has developed a number of coupled thermal-hydraulics/neutron-kinetics code systems which can solve the defined goal on assembly-wise or pin-by-pin level. Three types of couplings are applied depending on the utilized codes and application purposes: internal, external and parallel. An overview of the existing and future coupled systems available at GRS is given in this paper.
Development of multi-physics code systems based on the reactor dynamics code DYN3D
Authors: Kliem S. , Gommlich A. , Grahn A. , Rohde U. , Schütze J. , Frank T. , Gomez A. , Sanchez V.
Published: December 2013

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4.

The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D.
CANDU fuel elements behaviour in the load following tests
Authors: Horhoianu G. , Palleck S.
Published: December 2013

Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during test period. New LF tests are planed to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions.
A study on the damage of potential first wall materials in a nuclear fusion reactor using plutonium bearing salt
Authors: Übeyli M.
Published: December 2013

Selection of first wall material for fusion reactors is very crucial when taking into account of fusion blanket design and operation cost. A realistic way to extend the working period of first wall structure is to use a protective flowing liquid wall between fusion plasma and first wall. HYLIFE-II, one of the important fusion reactor design concepts, uses such a liquid wall. In the current article, the radiation damage on the first wall of HYLIFE-II fusion reactor was investigated for various candidate materials. In the liquid wall of the reactor, a molten salt containing weapon grade (WG) plutonium isotopes was used. The numerical results indicated that a refractory alloy of W-5Re was found to have the lowest damage values. In addition, the use of WG plutonium isotopes did not have a negative effect on the radiation damage characteristics of the investigated structural materials.
Preliminary evaluation of effect of Engineered Safety Features on source term for AHWR containment
Authors: Thangamani I. , Gera B. , Dutta A. , Verma V. , Singh R. K. , Ghosh A. K.
Published: December 2013

The proposed Advanced Heavy Water Reactor (AHWR) employs double containment envelope along with many Engineered Safety Features (ESFs) to mitigate the consequences of Loss-of-Coolant Accidents (LOCA) with safety system failure, during which high enthalpy steam and radioactive fission products will be discharged into the containment. In such conditions, the pressurized containment will be the source of activity release to the environment by way of leakage. It is required to study the effect of ESFs on the source term from the AHWR containment. An analysis was performed to evaluate the release rate from the AHWR containment during a postulated accident with the in-house containment code CONTRAN and the aerosol behavior code NAUA5-M in a coupled way. Modules for simulating the engineered safety features were incorporated in the CONTRAN code and the aerosol transport behaviour was evaluated using NAUA5-M separately. The AHWR containment is divided into three nodal volumes interconnected by junctions. The blow down mass, energy discharge data and activity released into the containment from the reactor core, for a postulated LOCA case of 200% RIH break with failure of shutdown systems (1 & 2), are inputs to the CONTRAN code. Thermodynamic parameters like containment gas temperature, partial pressure of steam, air in the subdivided volumes along with the flow rates through junctions obtained from CONTRAN were supplied to NAUA5-M. An analysis was carried out for a number of cases, postulated based on availability/unavailability of ESFs. Pressure, temperature and activity concentration transients were evaluated, for 72 h, in the subdivided volumes along with the activity released out of the containment through leakages and stack discharges for all the cases. This paper highlights the importance of operation of ESF in reducing the activity release to the environment.
Determination of 89Zr production parameters via different reactions using ALICE and TALYS codes
Authors: Taghilo M. , Kakavand T. , Sadeghi M.
Published: December 2013

89Zr is an important positron-emitting radionuclide for positron emission tomography (PET) and used in the field of tumor diagnostics, tumor therapy and the investigation of the bio-kinetic. The ALICE-91 and TALYS-1.0 codes were used to calculate excitation function for proton, alpha, deuteron and neutron induced on various targets that lead to the production of 89Zr radioisotopes using intermediate energy accelerators. Requisite thickness of the targets was obtained by SRIM code. The 89Zr production yield was evaluated using excitation function and stopping power. For the 89Y(p, n)89Zr reaction, the calculation data were compared with the experimental data. The 89Y(p, n)89Zr process was determined as most interesting one due to radionuclide purity. The ALICE-91 and TALYS-1.0 codes predict a maximum cross-section of about 939.86 mb at 15MeV and 860 mb at 13MeV respectively for this reaction.
Overview of safety improvement during RBMK-1500 reactor core lifetime upgrading
Authors: Pabarcius R. , Tonkunas A. , Slavickas A. , Seporaitis M.
Published: December 2013

Starting from early to mid 1970s the first generation of channel-type graphite-moderated boiling water reactors socalled RBMK types were designed and went into operation. Nowadays, all designed RBMKs represent three generations of reactors having significant differences with respect to their safety design features. The Ignalina NPP with two RBMK-1500 units being a second generation type were originally the most powerful reactors in the world. The change of important neutron-physical characteristics, being specific safety features for RBMK type reactors during RBMK-1500 core upgrading processes are analyzed until final shutdown and their impact on safety during reactor lifetime is discussed.
Training of the KBR nuclear emergency staff
Authors: Wink R. , Fary R.
Published: December 2013

In order to respond to a nuclear emergency case the first few hours are of decisive significance. For that reason sufficient and well-trained staff must be available at any time. The training of the personnel of the Nuclear power plant Brokdorf (KBR) involved in the emergency management organisation will be described in this presentation.
Preparation of emergency care centre exercises
Authors: Schnadt H. , Miska H.
Published: December 2013

Setup and operation of emergency care centres (sometimes also addressed as emergency reception centres) are part of emergency response in the environs of nuclear power plants. The preparation of an exercise scenario for such a centre is very demanding on the responsible agency. Therefore, a computer code has been developed which helps to translate the exercise objectives into instructions for figurants which simulate the affected population. These instructions are intended to steer a determined flow of people through the emergency care centre by providing fictitious radiological readings and by injecting the demand for additional actions of response personnel by statements and questions.
Independent assessor's report on the experience with preparation, managing, observing and evaluation of emergency exercises without prior notice in nuclear power plants
Authors: Matzig B. , Solisch T.-O.
Published: December 2013

We inspect regularly the measures taken in terms of the plants operator emergency management organisation in supposed emergency situations and assess their effectiveness on behalf of the supervisory authorities under the atomic energy law. To this end corresponding exercises based on scenarios that have been prepared by ESN are carried out in the plants and followed by a group of independent observers. The course and effectiveness of measures initiated are assessed in the form of an expert's report. ESN prepared, managed, observed und evaluated more than 75 emergency exercises in nuclear power plants in the last twenty years. With this experience and particularly with the experience feedback between plants we contributed to the development of the internal emergency management for nuclear power plants in Germany. This paper at hand summarizes some of our experience.
Planning, implementation and evaluation of an emergency exercise at the Paul Scherrer Institute
Authors: Hirt R. , Lörtscher Y.
Published: December 2013

Like all nuclear power plants in Switzerland, the Paul Scherrer Institute in Villigen (PSI), Switzerland, has the duty to conduct an emergency exercise every year. The two monitoring authorities ENSI (Swiss Federal Nuclear Safety Inspectorate) and the BAG (Federal Office of Public Health – FOPH) use this exercise for their yearly audit of the emergency organization. The emergency organization of the Paul Scherrer Institute consists of the emergency staff, the fire brigade, the first aid station, the traffic- and the security platoon and the radiation safety, the latter consisting of the decontamination group, the radio analytics group and the measurement group (for radioactivity dispersion calculations). The framework is set up by the ENSI beforehand, based on a fixed cycle. For the emergency exercise of 2009 a plane crashing into the interim storage facility of the federal government (BZL) at PSI was assumed. This contribution describes the approach of the planning process, the performance and the evaluation of the exercise.
Simulation of HALDEN IFA-650 loss-of-coolant accidents tests with TRACE
Authors: Aounallah Y.
Published: December 2013

The trends towards higher burnups, longer in-core fuel residence time, and the development of new cladding materials, have led to the need to re-assess the safety criteria for loss-of-coolant accidents. The HALDEN IFA-650 series of experiments are integral, single-pin, in-pile tests on the fuel behavior undergoing conditions of a loss-of-coolant accident over pre-determined fuel rod linear heat rate, and under thermally-controlled structural environment. Fresh and irradiated fuel pins from commercial reactors were tested. The present work seeks to assess the capability of a new (best-estimate) thermal-hydraulics code TRACE (version 5.0), to adequately predict the thermal-hydraulic impact of a loss-of-coolant accident on fuel rods. Several tests have been simulated over a range of fuel rod power and surrounding structure temperatures. The tests initiated from conditions of single-phase liquid natural circulation operating at 7MPa with subsequent venting to near-atmospheric pressure.
Upgrading (V)HTR fuel elements for generationIV goals by SiC encapsulation
Authors: Knorr J. , Kerber A. , Moormann R.
Published: December 2013

The pebble bed reactor is one of the most promising concepts for (V)HTR. Nevertheless recent re-evaluation of AVR and THTR results emphasizes once more that claimed advantages strongly depend on the properties of the fuel elements and their behavior under operational conditions. Additionally safeguards, waste management and disposal aspects gain increasing importance today. The conventional uncoated graphite pebbles meet only inadequately the requirements of Generation IV facilities. Since long experts agree, that corrosion-resistant pebbles with high retention capability for fission products would considerably improve the chances of the pebble bed reactor concept. With laser beam joining of ceramics the key technology is now available for the silicon carbide (SiC) encapsulation of (V)HTR components. The envisaged innovative fuel element consists of a robust SiC hollow sphere filled with moderator and TRISO coated particles. The positive assessment according to Gen IV criteria should justify necessary R&D efforts to obtain qualified fuel elements and demonstrate their superiority under operational conditions.
Slug modeling with 1D two-fluid model
Authors: Tiselj I. , Martin C. Samuel
Published: December 2013

Simulations of condensation-induced water hammer with one-dimensional two-fluid model requires explicit modeling of slug formation, slug propagation, and in some cases slug decay. Stratified flow correlations that are more or less well known in 1D two-fluid models, are crucial for accurate description of the initial phase of the slug formation and slug propagation. Slug formation means transition to other flow regime that requires different set of correlations. To use such two-fluid model for condensation induced water hammer simulations, a single slug must be explicitly recognized and captured. In the present work two cases of condensation-induced water hammer simulations performed with WAHA code, are described and discussed: injection of cold liquid into horizontal pipe filled with steam and injection of hot steam into horizontal pipe partially filled with cold liquid.
A discussion of hyperbolicity in CATHENA 4: Virtual Mass and phase-to-interface pressure differences
Authors: Aydemir N. U.
Published: December 2013

It is well known that the one-dimensional equations of motion for two-phase flow are non-hyperbolic. Non-hyperbolicity can lead to numerical instabilities, destroying the solution. However, researchers in the last few decades were able to show that inclusion of virtual mass and/or phase-to-interface pressure differences in the momentum equations successfully render the equations of motion hyperbolic. In the present paper, the effect of including virtual mass and phase-to-interface pressure terms in the momentum equations on the hyperbolicity of the two-phase model in the CATHENA 4 code is discussed. The study is motivated by the fact that the inclusion of either model has been shown in the open literature to lead to a hyperbolic system separately. However, no known study exists that examine hyperbolicity in the presence of both these terms in the momentum equations. In this work, both terms are considered in the model equations simultaneously and their implications on the hyperbolicity of the two-phase model are discussed. Specifically, it is shown that in the case of mixed flow, there is a distinct region of non-hyperbolicity that developers need to be aware of when their equations include both the virtual mass and the phase-to-interface terms. Selecting the coefficients of phase-to-interface pressure difference terms properly ensures that the equations are hyperbolic for a wide range of conditions.