bluuebunny/crossref_metadata_2025_split huggingface dataset:
CCF analysis for new reactor designs
Authors: Blombach J. , Bordihn S. , Kollasko H. 
Published: December 2013

Abstract
Common Cause Failures (CCF) have been an issue of safety consideration in nuclear safety since a long time. For the design of the European Pressurised Reactor (EPR) probabilistic safety goals in conjunction with the limit set by CCF on failure probability that can be achieved by redundancy led to the requirement to fulfil vital safety functions by functionally diverse systems. The probabilistic safety analysis (PSA) which was performed for EPR takes into account CCFs which are modelled using the beta-factor model of the European Utility Requirements (EUR). The PSA shows that the probabilistic safety goal for core damage frequency – CDF < 10−5/year – is met with sufficient margin. In addition to the modelling of CCF in the PSA a deterministic CCF analysis is being performed. It is a safety function based approach on component level analysing frontline and backup systems with respect to common components or type identical components. It ensures that diverse safety functions are fulfilled by systems being sufficiently diverse including all necessary support systems. Components are identified where further diversification or additional defences against CCF are required.
CCF analysis in PSA applications from a licensee view
Authors: Schubert B. 
Published: December 2013

Abstract
In highly redundant systems CCF dominate the calculated unavailabilities. CCF contribute over 90% to the resulting core damage frequencies of some PRAs for German Nuclear Power Plants. Because the results of PRAs can lead to costly back fitting measures, it is essential to ensure these results. The principles of CCF analysis and the efforts to establish a comprehensive data base are described. In addition the utilities have undertaken efforts to develop concepts for defensive strategies and to introduce these measures in the plant operation in order to reduce the CCF probabilities. Unfortunately these efforts are yet not fully supported by the existing models.
Multigroup neutron reflection and leakage conditions for axial core models with anisotropic scattering
Authors: de Abreu M. P. 
Published: December 2013

Abstract
In this article, multigroup discrete ordinates interface conditions for neutron reflection and leakage were derived for axial models of nuclear reactors with anisotropic scattering. These conditions are completely free from spatial truncation error. The key to the derivation is a suitable reformulation of exact discretized equations – the multigroup spectral nodal equations – defined in the nonactive boundary regions at top and bottom. The resulting interface conditions can be used to replace the nonactive boundary regions of a nuclear reactor core in axial reactor computations. In order to illustrate the improvements that can be experienced by most one-dimensional multigroup discrete ordinates codes with our multigroup interface conditions, we perform a numerical experiment with a representative multigroup code and an axial core model with anisotropic scattering.
Analyses of severe accident scenarios in RBMK-1500
Authors: Kaliatka A. , Rimkevicius S. , Uspuras E. , Urbonavicius E. 
Published: December 2013

Abstract
Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena.
Comparative assessment of methods for the reactivity measurement in subcritical systems by pulsed experiments
Authors: Canepa S. , Dulla S. , Ravetto P. 
Published: December 2013

Abstract
The determination of the subcriticality level for source-driven systems is an important problem that can be handled by several deterministic methods developed for the interpretation of flux signals in pulsed experiments. The paper compares the performance of several methods for different physical characteristics of the system considered and for different durations of the source pulse. The effects of the subcriticality level and of the experimental uncertainty are also investigated. Furthermore, an analysis of the influence of spatial effects is carried out. The study gives some indications on the most suitable method in a specific experimental configuration pointing out at the situations where non-conservative evaluations of the reactivity can be expected, i.e. when the system effective multiplication constant is underestimated.
Quantitative application of Monte Carlo simulation in Fire-PSA
Authors: Keski-Rahkonen O. , Mangs J. , Hostikka S. , Korhonen T. 
Published: December 2013

Abstract
In a power plant a fire cell forms the basic subunit. Since the fire is initially located there, the full-scale time dependent fire simulation and estimation of target response must be performed within the fire cell. Conditional, time dependent damage probabilities in a fire cell can now be calculated for arbitrary targets (component or a subsystem) combining probabilistic (Monte Carlo) and deterministic simulation. For the latter a spectrum from simple correlations up to latest computational fluid dynamics models is available. Selection of the code is made according to the requirements from the target cell. Although calculations are numerically heavy, it is now economically possible and feasible to carry out quantitative fire-PSA for a complete plant iteratively with the main PSA. From real applications examples are shown on assessment of fire spread possibility in a relay room, and potential of fire spread on cables in a tunnel.
Source term assessment as a basis for protective measures for the population in case of a nuclear accident in a nuclear power plant with radiological consequences
Authors: Zander N. , Löffler H. 
Published: December 2013

Abstract
The reliability of the evaluation of the radiological situation in the pre-release phase of a nuclear accident strongly depends on the quality of the source term estimation. Input data for the decision support system RODOS, such as time, composition, extent, and duration of a possible release of radioactive substances (source term) serve as a basis for the assessment of possible radiological consequences in the vicinity of a plant. With the collaboration of BfS and GRS a source term estimation tool was tested at a reference plant in order to provide data for RODOS.
Fast online system for forecasting environmental impact during an incident
Authors: Kümmerle E. , Möllmann-Coers M. , Pomplun E. , Hille R. 
Published: December 2013

Abstract
In an incident, the operational control staff needs a rapid general view of the radiological situation in the vicinity of the emission source which is suitable for decision support. Because the measuring team cannot provide sufficient measurements immediately, model calculations based on the available information about the course of the accident are additionally required. Such calculations do not only serve to create an overview of the current situation but can also provide prognoses for several hours. For this purpose, a software system for environmental impact forecasting was developed at Research Centre Jülich, which calculates time-integrated doses as well as ambient gamma dose rates and activity values. The latter can be directly compared with values measured during the incident and consequently can be used to assess the model assumptions.
Harnessing monitoring measurements in urban environments for decision making after nuclear accidents
Authors: Kaiser J. C. , Pröhl G. 
Published: December 2013

Abstract
This article gives an overview on the conceptual design of the Inhabited Areas Monitoring Module IAMM which will be introduced into European decision support systems for nuclear emergencies. It will improve the use of monitoring data of radioactive contamination in urban environments for decision making. IAMM converts the dated gamma dose rate (GDR) measurements from geo-referenced locations into maps of surface contamination with an enhanced spatial resolution. Depending on the availability of the monitoring data, IAMM relies on two modes of operation. If there are only a few measurements, these are taken to improve the maps from a deposition model using data assimilation. If the number of measurements is sufficient to apply spatial interpolation IAMM will rely entirely on monitoring data. Suitable geo-referenced data points will be interpreted by IAMM with respect to their detector environment using the concept of location factors. The endpoints of IAMM can be used directly for decision making or dose calculations with either simple dose models or the more refined EuRopean Model for INhabited areas (ERMIN).
Protection against dependent failures, analysis of dependencies and derivation of CCF data
Authors: Knochenhauer M. , Mankamo T. , Hellström P. , Johansson G. 
Published: December 2013

Abstract
The article describes a number of activities that were part of a major Nordic co-operation project which was recently performed, and which aimed at providing state-of-the-art guidance on the handling of dependent failures in nuclear power plants, both in operation and safety management, and as a part of plant PSA:s. This includes a rather wide scope of activities ranging from reviewing and describing available sources of data on CCF and CCF modelling techniques to the analysis of the efficiency of various protective measures against dependent failures in safety related equipment. A set of guidance documents was issued. The project also included a number of data analysis pilot applications, where CCF events in the ICDE database were analysed applying the impact vector method for several types of safety critical components.
Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
Authors: Vaidya A. M. , Maheshwari N. K. , Vijayan P. K. , Saha D. 
Published: December 2013

Abstract
Three dimensional CFD simulations of the moderator flow in the calandria vessel of a heavy water reactor are performed using the PHOENICS CFD code. The model includes the entire calandria vessel consisting of three shells, calandria tubes and inlet and outlet nozzle openings. The computational model prepared in PHOENICS consists of (a) standard k-∊ turbulence model, (b) PARSOL technique for handling curved objects in cartesian grids and (c) Boussinesq formulation for handling variable density flows. PHOENICS is validated by applying it to three different flow cases. The flow pattern in the calandria vessel under normal operating conditions is obtained through simulation. The effect of the presence of calandria tubes and heat generation on moderator flow pattern is studied. The simulation is also performed for various heat loads and moderator mass flow rates. The maximum temperature achieved by the moderator flow under various heat loads and moderator mass flow rates is obtained.
Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
Authors: Babilas E. , Urbonavicius E. , Rimkevicius S. , Kaliatka A. 
Published: December 2013

Abstract
The Accident Localisation System (ALS) of the Ignalina nuclear power plant is a “pressure suppression” type confinement, which protects the population, employees and environment from radiation hazards. According to the emergency operating procedures for the Ignalina nuclear power plant ∼110 m3 of hydrogen is released to the ALS compartments during the maximum design basis accident. However, in case of beyond design basis accidents, when the oxidation of zirconium starts, the amount of generated hydrogen could be significantly higher. The volume concentration of hydrogen in the compartment might reach 4 % of a combustible mixture. To prevent a possible hydrogen accumulation in the ALS of the Ignalina nuclear power plant during an accident, the H2 control system is installed. The analysis performed resulted in the identification of places of possible H2 accumulation in the ALS compartments during transient processes and the assessment of the mixture combustibility in these places for a postulated accident scenario. The results also allow to evaluate the influence of the proposed severe accident management strategy on the ALS conditions.
Notifiable work activities in water supply facilities – execution experience from Bavaria
Authors: Reifenhäuser C. , Körner S. 
Published: December 2013

Abstract
Since 2001, the protection against natural radiation is implemented in the German Radiation Protection Ordinance. Water supply facilities are now obliged to estimate the radon exposure of their staff. In 75 water supply facilities in Bavaria, Germany, radon exposures above the action limit of 6 mSv per year were determined. In these facilities the radon exposure has to be continuously monitored. By the proposal of the Bavarian Environment Agency (LfU), this should be done by continuous wearing of personal track-etch detectors, that are changed every three months. The results are sent to the LfU and reported to the central German dose register. Because of measuring periods of three months a high radon exposure that could probably exceed the annual limit of 20 mSv can be recognised at an early stage and immediate protection measures can be taken. In collaboration with the University of Regensburg remedial strategies for the permanent reduction of the radon exposure have been developed. The exposure is effectively reduced by reducing the time spent inside the units as much as possible. A very effective way to reduce the radon concentration inside buildings and hence the radon exposure is to blow fresh air directly into the work place of the staff with stationary or mobile devices. An alternative way to reduce the indoor air concentration is to avoid any transfer of radon polluted air exhausted from purification tanks to indoor air. For a long lasting success of the remedial measures a work instruction was developed by the LfU. All plants, their mean radon concentration and the measures to be taken have to be stated there.
Radiological evaluation of mine water discharges from hard coal mining in the Fossa Eugeniana Area, Germany
Authors: Wanke C. , Ritzel S. , Sachse R. , Michel R. 
Published: December 2013

Abstract
Discharges of radium-carrying mine waters from hard coal mining in the area of the Fossa Eugeniana and the Rheinberger Altrhein in Germany have caused elevated radioactivity in sediments and soils. Extensive measurements of the ambient dose rate revealed some small area contaminations as well as a larger one in a nature protection area at the Rheinberger Altrhein. For the evaluation of the potential radiation exposure investigations of the activity concentrations of river sediments, soils, and biological materials were performed. The potential exposure of members of the public was modelled on the basis of the German Guidelines for the Calculation of Radiation Exposure due to Radioactivity from Mining Activities. In the assessment the present situation as well as its possible future development was considered.
Consideration of statistical uncertainties for the determination of representative values of the specific activity of wastes
Authors: Barthel R. 
Published: December 2013

Abstract
The German Radiation Protection Commission has recommended “Principles and Methods for the Consideration of Statistical Uncertainties for the Determination of Representative Values of the Specific Activity of NORM wastes” concerning the proof of compliance with supervision limits or dose standards according to § 97 and § 98 of the Radiation Protection Ordinance, respectively. The recommendation comprises a method ensuring the representativeness of estimates for the specific activity of NORM wastes, which also assures the required evidence for conformity with respect to supervision limits or dose standards, respectively. On the basis of a sampling survey, confidence limits for expectation values of specific activities are determined, which will be used to show that the supervision limit or the dose standard is met or exceeded with certainty, or that the performed sampling is not sufficient for the intended assessment. The sampling effort depends on the type and the width of the distribution of specific activities and is determined by the position of the confidence interval with respect to the supervision limit or of the resulting doses with respect to the dose standard. The statistical uncertainties that are described by confidence limits may be reduced by an optimised extension of the sample number, as far as necessary.
Assessment of FEMAXI and TESPA-ROD codes for modelling of BDBA in RBMK-1500
Authors: Jusevičiu-tė A. , Kaliatka A. , Urbonavičius E. , Duškesas G. , Juodis L. , Sonnenburg H. G. 
Published: December 2013

Abstract
Processes occurring during the reactor operation and accidents change the thermal, chemical and other properties in nuclear fuel rods. This paper describes the processes in the nuclear fuel rods and tools for simulation at these processes. Fuel rods used at Ignalina NPP were simulated with FEMAXI-V (Japan) and TESPA-ROD (Germany) codes. Developed models were employed for the analysis of processes in fuel rods in case of large LOCA Beyond Design Basis Accident (BDBA). The developed models and results of the analysis are presented in this paper.
An analytical solution for the time-dependent SN transport equation in a slab
Authors: Segatto C. F. , Vilhena M. T. , Gonçalez T. T. 
Published: December 2013

Abstract
In this work an analytical solution for the time-dependent SN transport equation in a slab using the double Laplace transform technique is reported. This procedure leads to an analytical solution for the angular flux in integral form. Numerical simulations are presented and compared with results in the literature.
Design Concept of the High Performance Light Water Reactor
Authors: Schulenburg T. , Bittermann D. , Starflinger J. 
Published: December 2013

Abstract
The “High Performance Light Water Reactor” (HPLWR) is a Light Water Reactor operating with supercritical water as coolant. At a pressure of 25 MPa in the core, water is heated up from 280 to 500 °C. For these conditions, the envisaged net plant efficiency is 43.5 %. The core design concept is based on a so-called “3-pass-core” in which the coolant is heated up in three subsequent steps. After each step, the coolant is mixed avoiding hot streaks possibly leading to unacceptable wall temperatures. The design of such a core comprises fuel assemblies containing 40 fuel rods and an inner and outer box for a better neutron moderation. Nine of these are assembled to a cluster with common head- and foot piece. The coolant is mixed inside an upper and inside a lower mixing chamber and leaves the reactor pressure vessel through a co-axial pipe, which protects the vessel wall against too high temperatures.
Performance analyses of the communication networks of a modern supervision and control system of research reactors
Authors: El-Madbouly E. I. , Shaat M. K. , Shokr A. M. , Elrefaei G. H. 
Published: December 2013

Abstract
The functions of the Instrumentation and Control (I&C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated.
Optimizing the dynamic response of the H. B. Robinson nuclear plant using multiobjective particle swarm optimization
Authors: Elsays M. A. , Aly M. Naguib , Badawi A. A. 
Published: December 2013

Abstract
In this paper, the Particle Swarm Optimization (PSO) algorithm is modified to deal with Multiobjective Optimization Problems (MOPs). A mathematical model for predicting the dynamic response of the H. B. Robinson nuclear power plant, which represents an Initial Value Problem (IVP) of Ordinary Differential Equations (ODEs), is solved using Runge-Kutta formula. The resulted data values are represented as a system of nonlinear algebraic equations by interpolation schemes for data fitting. This system of fitted nonlinear algebraic equations represents a nonlinear multiobjective optimization problem. A Multiobjective Particle Swarm Optimizer (MOPSO) which is based on the Pareto optimality concept is developed and applied to maximize the above mentioned problem. Results show that MOPSO efficiently cope with the problem and finds multiple Pareto optimal solutions.
Thermal analytical model for MoO3 ampoules irradiated in a high neutron flux
Authors: Abd-El-Hady A. , Abou-El-Maaty T. 
Published: December 2013

Abstract
A one dimensional steady state thermal analytical model has been developed to study the heat transfer and temperature distributions in a quartz ampoule filled with MoO3 powder. The source heat generation inside the ampoule is released from the high neutron flux (1.4 × 1014 neutrons/cm2s) interaction with MoO3. Natural and forced convections heat transfer boundary conditions are adopted during the irradiation process. The peak temperatures in MoO3 powder and quartz are calculated and compared with their melting temperatures to ensure the irradiation safety criteria.
Integral analysis of low power and shutdown operation: insights and outcome
Authors: Spitzer C. 
Published: December 2013

Abstract
Probabilistic Safety Assessments (PSAs) in Germany have essentially been conducted in the framework of recurrent Safety Reviews (SRs). Besides, probabilistic investigations have also been submitted and reviewed in the usual regulatory procedures. Significant safety relevant insights have e. g. been gained from the Human Reliability Assessment (HRA) in the context of Low Power and Shutdown (LPS) PSAs whose performance has been launched by the regulatory body. In this paper major insights and results together with the practical outcome particularly from Human Factor (HF) and safety relevance points of view are described. Thereby, general key issues on HRA as well as substantial insights and conclusions derived from LPS HRAs are addressed. Further, a specific example with respect to the design and review of a complete set of operating procedures specifically dedicated to the shutdown operation is illustrated by outlining the general approach as well as the overall outcome achieved. Finally, conclusions are drawn regarding the relevance of probabilistic investigations, the appropriateness of a symptom oriented proceeding related to shutdown operation and the potential of utilising an advanced HF methodology.