1
Comparison of high purity fine grain graphites from different suppliers with regard to physical, mechanical and thermal properties : Final report. Als ms. Gedr. /
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2
Materials investigations for the SNQ Target Station : progress report 1985 /
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3
User's manual for ASTERIX-2 : a two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor /
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4
Graphitic matrix materials for spherical HTR fuel elements : results of material development and irradiation testing : catalogue of pictures and tables /
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5
Modelling of steam condensation in the primary flow channel of a gas heated steam generator /
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6
Mechanisms of surface scale formation in HTR helium /
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7
VSOP : computer code system for reactor physics and fuel cycle simulation /
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8
A model for irradiation induced changes in graphite material properties /
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Cords, H.
1978
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9
The role of solid-solution strengthening in the development of alloys for HTR applications /
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10
Contributions for the London International Conference on Carbon and Graphite. 5 : papers to be read at the conference in Imperial College, London, 18-22 September 1978 /
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11
Determination of equivalent cross sections for representation of control rod regions in diffusion calculations /
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12
A critical evaluation of the Mu-mode and Lambda-mode methods of evaluating reactor stability with respect to xenon-induced spatial power oscillations : design implications for the high temperature reactor /
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13
Theoretische Untersuchungen über die Thermodynamik einer Coated Particle, Grafitmatrix Feinstruktur /
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14
Numerical integration of the reactor kinetics equations with exponentially fitted implicit methods /
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15
A quick computational procedure for the assessment of xenon stability in HTR design studies /
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16
Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives /
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17
KPD : present worth fuel cycle calculation methods and codes /
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18
A computer controlled triple axis neutron spectrometer with CAMAC instrumentation /
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19
Wärmeübergang und Druckverlust an Reaktorbrennelementen in form längsdurchströmter Rundstabbündel /
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