Filtered generated 289 hits.
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2016:04 Uncertainty assessment of the SCANAIR V_7_5 computer program in analyses of BWR reactivity initiated accidents
In this work, we assess uncertainties that can be expected in computational analyses of boiling water reactor reactivity initiated accidents with the SCANAIR V_7_5 computer program. The work is intended to supplement a similar uncertainty assessment, focussed on pressurized water reactor accidents, which is currently being organized as an international benchmark exercise by the OECD Nuclear...
Content type: Publications -
2016:37 PARTRIDGE project: Review and evaluation of the probabilistic fracture mechanics code PRO-LOCA
Background The Swedish Radiation Safety Authority (SSM) and the Swedish nuclear power plant owners have financed Inspecta Technology in Sweden to evaluate the PRO-LOCA code. It is a computer code, developed by Battelle in the USA, in which the leak- and rupture probabilities of piping in nuclear power plants are analysed. The pipe systems may contain the damage mechanisms fatigue and/or...
Content type: Publications -
SKI Report 2008:01 Reliability Data for piping Components in Nordic Nuclear Poer Plants "R-Book" Project Phase 1
SKI Report 2008:01 - “Reliability Data for Piping Components in Nordic Nuclear Power Plants - R-Book Project Phase I” – is a planning document for a new R&D project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. Included in this handbook will be pipe leak failure rates and rupture frequencies that are...
Content type: Publications -
2010:16 Guidance to Risk-Informed Evaluation of Technical Specifications using PSA
No guidance documents for risk-informed analysis and assessment of changes in Limiting Conditions for Operation (LCO:s) in TS (Technical Specifications) have been available in the Nordic countries. In the U.S. several guidance documents, primarily developed by the owners groups for PWRs and BWRs, exist. Guidance for analysis of TS changes in this report generally follows the U.S. Regulatory...
Content type: Publications -
2010:42 Using the EPRI Risk-Informed ISI Methodology on Piping Systems in Forsmark 3
In the SSM regulation SSMFS 2008:13, it is stated that the selection of locations for inspection of mechanical components shall be based upon the risk for core damage or the risk of release of radioactive substances. Both qualitative and quantitative measures of the relative risk are allowed to be used. So far only the PWRs in Sweden are using a quantitative procedure to evaluate the risk...
Content type: Publications -
2009:07 Dependency Analysis Guidance Nordic/German Working Group on Common Cause Failure analysis
The Regulatory Code SSMFS 2008:1 of Swedish Radiation Safety Authority (SSM) includes requirements regarding the performance of probabilistic safety assessments (PSA), as well as PSA activities in general. Therefore, the follow-up of these activities is part of the inspection tasks of SSM. According to the SSMFS 2008:1, the safety analyses shall be based on a systematic identification and...
Content type: Publications -
2014:18 Evaluation of the Halden IFA-650 loss-of-coolant accident experiments 2, 3 and 4
The analytical tools, which are fuel rod computer codes, that Quantum Technologies AB use and develop, contain models of several of the phenomena that are acting on the nuclear fuel (cladding temperature, fission gas driven pressure, strain and stress in the cladding, rod rupture, etc.) and how the separate effects interact in the complex integrated manner. The codes are under constant...
Content type: Publications -
2009:27 Analysis Strategy for Fracture Assessment of Defects in Ductile Materials
SSM has supported research for investigating the role of secondary stresses when fracture assessments are performed for cracked structures made of ductile materials. There are evidences that indicate that some secondary stresses, such as weld residual stresses, are not as important as primary stresses for estimating the safety margin against rupture (measured by the J-integral) for the type...
Content type: Publications -
2010:46 Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report
The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a...
Content type: Publications -
2007:15 Evaluation of the FRAPTRAN -1.3 Computer Cod
The FRAPTRAN-1.3 computer code has been evaluated regarding its applicability, modelling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under reactor power and coolant transients, such as overpower transients, reactivity initiated accidents (RIA),...
Content type: Publications