Forsmarks Kraftgrupp AB

Osthammar, Sweden

Forsmarks Kraftgrupp AB

Osthammar, Sweden
SEARCH FILTERS
Time filter
Source Type

Amft M.,KTH Royal Institute of Technology | Amft M.,Forsmarks Kraftgrupp AB | Walle L.E.,Norwegian University of Science and Technology | Ragazzon D.,Uppsala University | And 5 more authors.
Journal of Physical Chemistry C | Year: 2013

We show that the formation of the wetting layer and the experimentally observed continuous shift of the H2O-OH balance toward molecular water at increasing coverage on a TiO2(110) surface can be rationalized on a molecular level. The mechanism is based on the initial formation of stable hydroxyl pairs, a repulsive interaction between these pairs, and an attractive interaction with respect to water molecules. The experimental data are obtained by synchrotron radiation photoelectron spectroscopy and interpreted with the aid of density functional theory calculations and Monte Carlo simulations. © 2013 American Chemical Society.


Anglart H.,KTH Royal Institute of Technology | Alavyoon F.,Forsmarks Kraftgrupp AB | Novarini R.,KTH Royal Institute of Technology
Nuclear Engineering and Design | Year: 2010

The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate. © 2008 Elsevier B.V. All rights reserved.


Tinoco H.,Forsmarks Kraftgrupp AB | Buchwald P.,KTH Royal Institute of Technology | Frid W.,KTH Royal Institute of Technology
Nuclear Engineering and Design | Year: 2010

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several minutes. © 2008 Elsevier B.V. All rights reserved.


Grant
Agency: European Commission | Branch: FP7 | Program: CSA-CA | Phase: Fission-2007-2.1-03 | Award Amount: 2.11M | Year: 2008

The objective of this coordination action is to develop best practice guidelines for the performance of Level-2 PSA methodologies with a view to harmonization at EU level and allowing a meaningful and practical uncertainty evaluation in a Level-2 PSA. Speficic relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, research or services companies) will be established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. A technical group of level 2 PSA experts from project partners will propose guidelines for a limited-scope and a full-scope L2 PSA based on practical experience of each partner. Elaboration of this guidelines for operating plants will be the main activity of this coordination action. This guideline will be discussed with End-Users community and their opinion will be taken into account in the final version. The applicability of such a guideline for future reactor (Gen IV) will be also examined.


Marcinkiewicz J.,Forsmarks Kraftgrupp AB | Taler J.,Cracow University of Technology | Cebula A.,Cracow University of Technology
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013

The presented work is a result of actions taken in connection to analyses of root cause of damages (cracks and one brake) in the control rod shafts in Swedish BWR Forsmark 3. The damages were detected during the refueling outage 2008. It has been found that damages were caused by thermal fatigue. Extensive analyses of flow and temperature fields around the shaft were performed using transient CFD calculations [1, 2]. The character of the fluctuating thermal loadings on the shaft was confirmed by a limited experiment [3]. However the CFDcalculations of heat transfer between the water and the shaft have not been validated experimentally. In order to validate CFD-calculations of heat transfer between the water and the solid body the measurements of the non-stationary heat transfer are planned. The paper presents the method of determination of heat flux and temperature on the surface of the body based on temperature measurements at some discrete points beneath the surface and solving the inverse heat conduction problem (IHCP). Software was developed for performing measurements and calculations. Main parts of measurement system particularly design and manufacturing of measuring items, thermocouple installation, construction of test stand for initial testing and calibration are described. Verification of thermocouple locations was performed using computer tomography and the actual locations were introduced in to the calculation algorithm in order to improve accuracy of heat flux determination. Heat flux measuring error has been determined based on assumed random error in temperature measurement and accuracy of location verification (computer tomography). Results of initial verifying tests are presented and discussed. The measuring system is now ready for performing measurements of transient heat transfer in configurations that can occur in a reactor environment. Copyright © 2013 by ASME.


Grant
Agency: European Commission | Branch: FP7 | Program: CP-FP | Phase: Fission-2010-2.1.1 | Award Amount: 3.96M | Year: 2011

The assessment of the condition of low-voltage instrumentation, control, and power cables in nuclear power plants is of increasing importance as plants age and lifetime extensions are envisaged. Furthermore as new reactors are being constructed and many other are planned for the near future, the initial cables choice and the use of effective in-situ condition monitoring (CM) techniques to follow cable condition indicators from the beginning, can result to be very valuable at a later time for an effective cable lifetime management. The overall objective of the proposed project is to adapt, optimise and assess electrical CM techniques for nuclear cables that would allow utilities to assess in-situ the current cable degradation condition and, together with the establishment of appropriate acceptance criteria, to verify its qualified state over its entire length and to estimate its residual lifetime. To this extent, the project will consist in studying with accelerated ageing tests a representative selection of cables already installed in European Nuclear Power Plants (NPPs) in order to evaluate the ability of electrical CM techniques to detect local and global cable ageing. The results will be compared and correlated to those obtained with more conventional CM techniques for validation and residual life estimation. These tests will be supported by the study of the impact of cable polymers ageing on the electrical properties. These studies will allow not only to guide the adaptation and the optimisation of existing CM techniques, but also to interpret the results of the electrical measurements, to extend the validity of the results to other similar cables and to adapt the future cable design and formulations to electrical CM techniques. This investigation on innovative cables for future plants could open the way to a new generation of intelligent cables with improved diagnostic capability.


Masuda Y.,Japan National Institute of Advanced Industrial Science and Technology | Yoneya M.,Japan National Institute of Advanced Industrial Science and Technology | Suzuki A.,Japan National Institute of Advanced Industrial Science and Technology | Kimura S.,Kanazawa University | Alavyoon F.,Forsmarks Kraftgrupp AB
International Communications in Heat and Mass Transfer | Year: 2010

Two peculiar convection patterns-re-oscillation and stable non-centrosymmetric convection-are observed when two-dimensional double-diffusive convection in a porous enclosure (aspect ratio = 1.5) is analysed numerically. The top and bottom walls of the enclosure are insulated; constant and opposing heat and mass fluxes are prescribed on the vertical walls. Re-oscillation occurs when the convection pattern changes from centrosymmetric to non-centrosymmetric. When the buoyancy ratio, which generates re-oscillation convection, is marginally lower, the convection pattern changes to stable non-centrosymmetric. These two convection patterns can be observed only for limited values of the Rayleigh number, Lewis number, and buoyancy ratio. © 2009 Elsevier Ltd. All rights reserved.


Karlsson A.,Forsmarks Kraftgrupp AB | Frisk M.,Risk Pilot AB | Hultqvist G.,Havsbrus Consulting
PSAM 2014 - Probabilistic Safety Assessment and Management | Year: 2014

Benchmarking is an important activity in order to eliminate unjustified differences between PSA models and enable harmonisation. It could also be used in order to understand plant differences. As part of the BWR-club PSA activities benchmarking of bottom LOCA during outage, reactor level measurement and dominating initiating events have been performed. Modelling of bottom LOCA during outage varies between the BWR-club members and work performed within the BWR-club aims at compiling and understanding these differences. When it comes to reactor level measurement modelling varies from a more detailed modelling to more of a "black box" approach. Information has also been collected from the BWR-club members regarding dominating initiating events in their PSA studies. The initiating event frequencies, scope of the PSA studies and risk importance of different initiating events vary between the BWR-club members and work has been performed compiling and understanding these differences. BWR-club reports have been issued for bottom LOCA during outage and reactor level measurement, while the benchmarking of dominating initiating events is yet to be finalised.


Karlsson A.,Forsmarks Kraftgrupp AB | Hultqvist G.,Forsmarks Kraftgrupp AB | Frisk M.,Risk Pilot AB
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, PSA 2013 | Year: 2013

Benchmarking is an important activity in order to eliminate unjustified differences between PSA models and enable harmonisation. It could also be used in order to understand plant differences. As part of the BWR-club PSA activities benchmarking of bottom LOCA during outage, reactor level measurement and dominating initiating events is performed. The modelling of bottom LOCA during outage varies between the BWR-club members and work performed within the BWR-club aims at compiling and understanding these differences. BWR-club members have also filled in a questionnaire considering reactor level measurements. In general reactor level measurement could be modelled more in detail or more as a black box. Information is also collected from the BWR-club members regarding dominating initiating events in their PSA studies. The initiating event frequencies, scope of the PSA studies and risk importance of different initiating events vary between the BWR-club members and work will be performed compiling and understanding these differences. During the spring and summer of 2013 the results will be compiled into BWR-club reports.


Hultqvist G.,Forsmarks Kraftgrupp AB
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011 | Year: 2011

The Nordic PSA Group NPSAG was founded in December 2000 by the nuclear utilities in Finland and Sweden. In addition, the Swedish Nuclear Power Inspectorate (SKI) participates as an observer, and also takes part in the funding of many of the projects. NPSAG is intended to be a common forum for discussion of issues related to probabilistic safety assessment (PSA) of nuclear power plants, with focus on research and development needs. The group follows and discusses current issues related to PSA nationally and internationally, as well as PSA activities at the participating utilities. The group initiates and co-ordinates research and development activities and discusses how new knowledge shall be used. Important on-going activities concern CCF and dependent failures in general, as well as applications of PSA. In addition, a general and quite extensive discussion has been initiated about data for PSA models. The discussion concerns a number of issues, ranging from types of data needed to future procedures for data collection, processing and analysis. Over the years, international contacts have increased, especially with partners in Europe (initiated by BWROG Associate program and EU-research contacts). This is in line with the group's aim to create a common and lasting basis for the performance of PSA and for risk informed applications of PSA in Europe. One important result is a common pilot project with VGB (Germany) on multi-national CCF data analysis. The paper gives an overview of NPSAG projects - past and present, and of the types of international contacts and information collection activities of the group.

Loading Forsmarks Kraftgrupp AB collaborators
Loading Forsmarks Kraftgrupp AB collaborators