Table Of Content

Nuclear ig Engineering ier Sth and Design EVIER Nuclear Engineering and Design 171 (1997) 189-190 CONTENTS OF ISSUE 171 (1-3) Extended and Updated Selected Papers from the 20th MPA Seminar, Stuttgart, Germany, 6—7 October, 1994 Preface US Nuclear Regulatory Commission research for primary system integrity regulations C.Z. Serpan, Jr., M.E. Mayfield and J. Muscara Evaluation of piping damage in German nuclear power plants K.W. Bieniussa and H. Reck Evaluation of ductile tearing in cracked pipes and elbows under bending D. Moulin, G. Clement, B. Drubay and G. Goudet Manufacturing and material properties of forgings for the reactor pressure vessel of the high temperature engineering test reactor I. Sato and K. Suzuki The basic concept to ensure the integrity of components during operation W.G. Hienstorfer, J. Bartonicek and H. Kockelmann Replacement of safety-related valves in Philippsburg | nuclear power plant H. Wolf Tests for the dynamic behaviour of isolation valve cone screws K.-D. Tulke, W. Stoppler and G. Stern Nondestructive characterization of materials: A growing demand for describing damage and service-life-relevant aging processes in plant components G. Dobmann, N. Meyendorf and E. Schneider Cracking in stabilized austenitic stainless steel piping of German boiling water reactors—character- istic features and root cause M. Erve, U. Wesseling, R. Kilian, R. Hardt, G. Briimmer, V. Maier and U. Ilg Planned and implemented remedial measures to prevent IGSCC and to ensure safe plant operation with BWR piping systems made of stabilized austenitic steels M. Erve, G. Briimmer, U. Kleen, V. Maier, U. Ilg, H.J. Baumler, A. Seibold and D. Blind Assessment of aging of Zr—2.5Nb pressure tubes in CANDU™ reactors M.-P. Puls Essential results of analyses accompanying the leak rate experiments E22 at HDR H. Grebner, A. Hofler and H. Hunger Investigations into the ratchetting behaviour of austenitic pipes D. Kramer, S. Krolop, A. Scheffold and R. Stegmeyer 190 Contents of Volume/Nuclear Engineering and Design 171 (1997) 189-190 Analysis of Novo-Voronezh NPP unit 3 radiation lifetime of the WWER-440 reactor pressure vessel given samplings V. Kiselyov, A. Korinets, Y. Maksimov and V. Piminov Effect of cladding on the initiation behaviour of finite length cracks in an RPV under thermal shock J.G. Blauel, L. Hodulak, G. Nagel, W. Schmitt and D. Siegele Contents of Volume 171 Author Index of Volume 171 Subject Index of Volume 171 Instructions to Contributors bier