Nuclear Engineering and Design

Title Publication Date Language Citations
On modelling of kinematic hardening for ratcheting behaviour1995/01/01
Editorial2000/04/01English
Measurement of large negative reactivity in WWER-440. Sources of uncertainty1998/03/01English
Ratcheting deformation of advanced 316 steel under creep–plasticity condition1999/10/01English
Natural circulation flow behavior at reduced inventory in a VVER geometry2002/06/01English
Subject Index of Volume 2152002/06/01English
Risk-informed inservice inspection program2000/02/01English
Low alloy steel piping test for fracture criteria of leak before break1999/07/01English
Stresses in ellipsoidal pressure vessel heads with noncentral nozzle2000/06/01English
Some features of the nuclear heating reactor (NHR) design in China1995/05/01English
A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility2000/05/01English
Aspects on decommissioning of the Greifswald nuclear power plant1995/11/01English
Modified solution of radial nozzle with thick reinforcement in spherical vessel head subjected to radial load1997/12/01English
Study on reactor building structure using ultrahigh strength materials1995/06/01English
Load-carrying capacity and crack resistance of a cladding by the Sigma-oscillating wire technique1999/06/01English
The properties of WWER-440 type reactor pressure vessel steels cut out from operated units2000/02/01English
Evaluation of mechanical behavior of new type bellows with two-directional convolutions2000/04/01English
Fatigue monitoring and life cycle analysis of a recuperative heat exchanger in the primary circuit1995/10/01English
Vocational training related ENEN activities and their impact on nuclear competence development in Europe2024/04/01English
Thermal-mechanical safety analysis of heat pipe micro reactor2024/04/01English
Repair and replacement of reactor internals for plant life extension1998/10/01English
Improved Analysis of the Short-Term Station Blackout Accidents of the Peach Bottom Unit-2 Reactor with ASTEC Including Radiological Impact and Statistical Analysis with JRODOS2024/04/01English
The frequency effect on the fatigue crack growth rate of 304 stainless steel1999/07/01English
Scaling in the safety of next generation reactors1998/11/01English
Determination of limits for smallest detectable and largest subcritical leakage cracks in piping systems1995/10/01English
Fluid-structure interaction in non-rigid pipeline systems1997/07/01English
Remaining life management systems: from stand-alone to corporate memory systems and Internet (ALIAS system of MPA Stuttgart)1999/05/01English
Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation; material: 15 Ch2MFA ≈0.15 C–2.5 Cr–0.7 Mo–0.3 V1999/10/01English
Tests on large scale LMFR piping Part I: crack resistance properties of through-cracked straight 700 mm nominal diameter pipes tested under bending at ambient temperature and 550°C1995/09/01English
Damage mechanics analysis (Gurson model) and experimental verification of the behaviour of a crack in a weld-cladded component1997/10/01English