Soanes, T.P.T.
Central Electricity Generating Board, Berkeley (UK). Berkeley Nuclear Labs1980
Central Electricity Generating Board, Berkeley (UK). Berkeley Nuclear Labs1980
AbstractAbstract
[en] COMULIP (COre Movements Using LInear Programming) is a set of computer programs which allows the user to find the maximum movements of bricks in a layer of 16 sided polygonal bricks and 4 sided interstitial bricks, as in the core of the Hinkley 'B' Advanced Gas Cooled Reactor. The user supplies the dimensions of the bricks and keys, the boundary constraints and specifies the brick and direction for which the maximum movement is to be found. To facilitate data input, a standard set of dimensions is entered together with individual brick dimensions which differ. The linear programming method is used to maximise an objective function, which in this case is the movement of a specified brick in a given direction, subject to a set of linear constraints, which define the geometrical condition that no brick should overlap. The elasticity of the bricks is not taken into account, the effect of this being small compared with that of the clearances. The model analysed is a two dimensional horizontal cross-section through the core. This report forms the user's guide to the computer programs and gives details of the input data required and the output produced, as well as a brief description of the solution technique. The program may be used to study the effect of dimensional changes and brick or key damage, which occur during the lifetime of the reactor, on the movement of bricks. (author)
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Jul 1980; 46 p
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Report
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Holt, P.J.; Soanes, T.P.T.
The 9th international symposium on the packaging and transportation of radioactive materials. Proceedings: Volume 11989
The 9th international symposium on the packaging and transportation of radioactive materials. Proceedings: Volume 11989
AbstractAbstract
[en] Irradiated fuel transport flasks are designed to survive impact and fire accidents without significant leakage of contents. The impact standards laid down by the IAEA include a drop test from a height of 9 meters, in the most damaging attitude, onto a flat unyielding target. This requirement leads to the incorporation of energy absorption features for impact protection. Such features may be integral with the flask body and/or lid, as seen on the CEGB's Magnox MkM2 and AGR MkA2 flasks. Alternatively, it may be operationally more convenient to have removable energy absorbers, as on most LWR flasks and the CEGB's AGR MkA1 flask. This paper is concerned with the design of removable energy absorbers and the materials which may be used in their construction. In order to address the design issues, a notional cylindrical flask of 10 ton mass is considered. To allow a design margin, and to cater for all future requirements and possible changes in regulations etc., the drop height considered is increased to 36 meters
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Source
Oak Ridge National Lab., TN (USA); 492 p; 1989; p. 71-78; International symposium on packaging and transportation of radioactive materials; Washington, DC (USA); 11-16 Jun 1989; NTIS, PC A22/MF A01 as DE90004447
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Report
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Conference
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Fullard, K.; Dowler, H.J.; Soanes, T.P.T.
The resistance to impact of spent Magnox fuel transport flasks. Papers presented at a seminar held in London on 30 April and 1 May 19851985
The resistance to impact of spent Magnox fuel transport flasks. Papers presented at a seminar held in London on 30 April and 1 May 19851985
AbstractAbstract
[en] A 60mph impact into a tunnel abutment, of a flask on a railway flatrol with following vehicles, is shown to be a much less severe event for the flask than a 9 metre drop test to IAEA regulations. This involves the use of mathematical models of the full scale event of the same type as were employed in studying the behaviour of quarter scale models. The latter were subject to actual impact testing as part of the validation process. (author)
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Institution of Mechanical Engineers, London (UK); British Nuclear Energy Society, London; 249 p; ISBN 0 85298 574 6; ; 1985; p. 149-156; Mechanical Engineering Publications Ltd; Bury St. Edmunds (UK); Seminar on the resistance to impact of spent Magnox fuel transport flasks; London (UK); 30 Apr - 1 May 1985
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Book
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Conference
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Soanes, T.P.T.; Lewis, D.J.
Gas-cooled reactors today. (Volume 4). Papers, discussion, closing address, corrigenda. Proceedings of the conference held in Bristol on 20-24 September 19821983
Gas-cooled reactors today. (Volume 4). Papers, discussion, closing address, corrigenda. Proceedings of the conference held in Bristol on 20-24 September 19821983
AbstractAbstract
[en] The BERSAFE finite element system has been used to examine the structural integrity of graphite core bricks. Internally produced stresses due to irradiation shrinkage and creep have been examined using a special graphite version of the program. The effect of different corrosion profiles on the load bearing capacity of a brick has also been examined. The existing approximate rule for determining brick strength using the concept of an effective weightloss has been evaluated. (author)
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British Nuclear Energy Society, London; 328 p; ISBN 0 7277 0168 1; ; 1983; p. 137-142; BNES; London (UK); Gas-cooled reactors today; Bristol (UK); 20-24 Sep 1982
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Book
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Conference
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INIS VolumeINIS Volume
INIS IssueINIS Issue
Soanes, T.P.T.; Lewis, D.J.
Gas-cooled reactors today. (Vol.4). Papers. Discussion. Closing address. Corrigenda1983
Gas-cooled reactors today. (Vol.4). Papers. Discussion. Closing address. Corrigenda1983
AbstractAbstract
[en] The BERSAFE finite element system has been used to examine the structural integrity of graphite core bricks. Internally produced stresses due to irradiation shrinkage and creep have been examined using a special graphite version of the program. The effect of different corrosion profiles on the load bearing capacity of a brick has also been examined. The existing approximate rule for determining brick strength using the concept of an effective weightloss has been evaluated. (author)
Primary Subject
Source
British Nuclear Energy Society, London; p. 137-142; ISBN 07277 0169 X (COMPLETE SET); ; 1983; p. 137-142; British Nuclear Energy Society; London (UK); Symposium on gas-cooled reactors today; Bristol (UK); 20-24 Sep 1982; Available from B.N.E.S., 1-7 Great George St., London SW1P 3AA
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Book
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CARBON, CARBON COMPOUNDS, CARBON OXIDES, CHALCOGENIDES, CHEMICAL REACTIONS, COMPUTER CODES, ELEMENTS, ENRICHED URANIUM REACTORS, GAS COOLED REACTORS, GRAPHITE MODERATED REACTORS, MECHANICAL PROPERTIES, NATURAL URANIUM REACTORS, NONMETALS, NUMERICAL SOLUTION, OXIDES, OXYGEN COMPOUNDS, POWER REACTORS, REACTORS
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Dowler, H.J.; Soanes, T.P.T.
The resistance to impact of spent Magnox fuel transport flasks. Papers presented at a seminar held in London on 30 April and 1 May 19851985
The resistance to impact of spent Magnox fuel transport flasks. Papers presented at a seminar held in London on 30 April and 1 May 19851985
AbstractAbstract
[en] To assess the behaviour of a full scale flask and flatrol during a proposed demonstration impact into a tunnel abutment, a mathematical modelling technique was developed and validated. The work was performed at quarter scale and comprised of both scale model tests and mathematical analysis in one and two dimensions. Good agreement between model test results of the 26.8m/s (60 mph) abutment impacts and the mathematical analysis, validated the modelling techniques. The modelling method may be used with confidence to predict the outcome of the proposed full scale demonstration. (author)
Primary Subject
Source
Institution of Mechanical Engineers, London (UK); British Nuclear Energy Society, London; 249 p; ISBN 0 85298 574 6; ; 1985; p. 135-147; Mechanical Engineering Publications Ltd; Bury St. Edmunds (UK); Seminar on the resistance to impact of spent Magnox fuel transport flasks; London (UK); 30 Apr - 1 May 1985
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Book
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Conference
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Soanes, T.P.T.; Bell, W.; Vibert, A.J., E-mail: timothy.soanes@british-energy.com2005
AbstractAbstract
[en] Following the discovery of incorrect weld metal in the steam side shell to tubeplate weld in a type 316H stainless steel superheater steam header, a repair strategy had to be determined. The strategy adopted was to remove the incorrect weld material, which extended around the full circumference, by machining from the inside of the header, followed by rewelding from the inside using an automatic welding process and localised post-weld heat treatment. Due to concern over potential reheat cracking of the repair after return to service, a considerable amount of residual stress modelling was carried out to support the development and optimisation of a successful repair and heat treatment strategy and thus underwrite the safety case for return to service
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S0308-0161(04)00200-5; Copyright (c) 2004 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
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Journal Article
Journal
International Journal of Pressure Vessels and Piping; ISSN 0308-0161; ; CODEN PRVPAS; v. 82(4); p. 311-318
Country of publication
ALLOYS, AUSTENITIC STEELS, CALCULATION METHODS, CARBON ADDITIONS, CHROMIUM ALLOYS, CHROMIUM STEELS, CHROMIUM-MOLYBDENUM STEELS, CHROMIUM-NICKEL STEELS, CHROMIUM-NICKEL-MOLYBDENUM STEELS, CONTAINERS, CORROSION RESISTANT ALLOYS, ELEMENTS, FABRICATION, HEAT RESISTANT MATERIALS, HEAT RESISTING ALLOYS, HIGH ALLOY STEELS, IRON ALLOYS, IRON BASE ALLOYS, JOINING, JOINTS, MATERIALS, MATHEMATICAL SOLUTIONS, MECHANICAL PROPERTIES, MOLYBDENUM ALLOYS, NICKEL ALLOYS, NUCLEAR FACILITIES, NUMERICAL SOLUTION, POWER PLANTS, STAINLESS STEELS, STEEL-CR17NI12MO3, STEELS, STRESSES, THERMAL POWER PLANTS, TRANSITION ELEMENT ALLOYS
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