Filtered generated 396 hits.
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2007:10 DECOVALEX-THMC - Task D
The DECOVALEX project is an international cooperative project initiated by SKI, the Swedish Nuclear Power Inspectorate, with participation of about 10 international organizations. The name DECOVALEX stands for DEvelopment of COupled models and their VALidation against Experiments. The general goal of this project is to encourage multidisciplinary interactive and cooperative research on...
Content type: Publications -
2008:23 SKI’s and SSI’s review of SKB’s safety report SR-Can
This report summarises the Swedish Nuclear Power Inspectorate’s (SKI) and the Swedish Radiation Protection Authority’s (SSI) joint review of the Swedish Nuclear Fuel & Waste Management Co’s (SKB) safety report SR-Can, on post-closure safety for a KBS-3 spent nuclear fuel repository at Forsmark and Laxemar respectively (SKB TR-06-09). As of 1 July 2008 the Swedish Radiation Safety...
Content type: Publications -
2007:32 Low pH Cements
Concrete and cement are used in constructions as well as in conditioning of waste in repositories for radioactive waste. The development of low pH cements for use in geological repositories for radioactive waste stem from concerns over the potential for deleterious effects upon the host rock and other EBS materials (notably bentonite) under the hyperalkaline conditions (pH > 12) of cement...
Content type: Publications -
International peer review of repository application
The Swedish Radiation Safety Authority has performed a review of SKB’s (i.e. Svensk Kärnbränslehantering AB) application for construction of a repository for spent nuclear fuel, and recommends approval of this application, as stated in our pronouncement to the Government on 23 January 2018. A peer review has also been performed by OECD’s Nuclear Energy Agency (NEA) concerning the...
Content type: Regular Pages -
2017:11 Extended Common Load Model: A tool for dependent failure modelling in highly redundant structures
Background The treatment of dependent failures is one of the most controversial subjects in reliability and risk analyses. The difficulties are specially underlined in the case of highly redundant systems, when the number of redundant components or trains exceeds four. The Common Load Model (CLM), originally defined in the 70’ies, differs from other CCF models, as it relies on a specific...
Content type: Publications -
2008:18 Concerns when designing a safeguards approach for the back-end of the Swedish nuclear fuel cycle
Sweden has for many years collected the spent nuclear fuel originating from nuclear power plants. This fuel must at all times be kept under supervision to render a diversion impossible; this is of course due to the possibility to make weapons from the material. One idea is to keep the nuclear material in a repository deep under the ground; this is not only to keep the material safe from theft...
Content type: Publications -
2013:08 Licensing of safety critical software for nuclear reactors
It is widely accepted that the assessment of software cannot be limited to verification and testing of the end product, i.e. the computer code. Other factors such as the quality of the processes and methods for specifying, designing and coding have an important impact on the implementation. Existing standards provide limited guidance on the regulatory and safety assessment of these factors.
Content type: Publications -
2011:16 Modelling of ultrasonic testing of cracks in cladding
During the last two decades, SSM has supported research to develop a model for the non-destructive test situation based on ultrasonic technique. Such a model is important in many ways, for example to supplement and plan experimental studies and to perform parametric studies in qualification situation. Modeling can be a useful tool when the inspection system shall be technically justified.
Content type: Publications -
2013:24 Modelling of nuclear fuel cladding under loss-of-coolant accident conditions
We present a unified model for calculation of zirconium alloy fuel cladding rupture during a postulated loss-of-coolant accident in light water reactors. The model treats the Zr alloy solid-to-solid phase transformation kinetics, cladding creep deformation, oxidation and rupture as a function of temperature and time in an integrated fashion during the transient. The fuel cladding material...
Content type: Publications -
2011:04 Evaluation of the Technical Basis for New Proposals of Fatigue Design of Nuclear Components
During the recent years fatigue analysis procedures for nuclear components have been investigated. The most common method so far has been the American code ASME III. The basis for the current design procedures in ASME III is quite old and has now been evaluated against modern data leading to the proposal of modified design curves. Also the effect of the environment has been the subject of...
Content type: Publications