Steam Generator Ageing Management in Slovakia current practices and related issues

Size: px
Start display at page:

Download "Steam Generator Ageing Management in Slovakia current practices and related issues"

Transcription

1 International Atomic Energy Agency Meeting on Update and Upgrade of IAEA-TECDOC on Ageing Management of Steam Generators IAEA, Vienna, June 2009 Steam Generator Ageing Management in Slovakia current practices and related issues Martin Březina

2 Content of the presentation Introduction of VUJE company Basic description of SG WWER 440 SG in Slovak republic Operational problems and other issues Technical upgrades of SG WWER 440 Ageing management programs Conclusions

3 Introduction of the company VUJE a.s. An engineering company that performs design, supply, implementation, research and training activities, particularly in the field of nuclear and conventional power generation

4 Introduction of the company The company performs all activities related to the design, construction, operation, reconstruction and decommissioning of the following energy installations and systems: nuclear power plants hydro-electric power plants, fossil-fuel power plants and heating plants with coal, gas and oil used as fuel power plants and heating plants for waste wood combustion wind power plants high-voltage transmission lines kv high-voltage distribution plants development and supply of simulators used for training the operating personnel of power plants, distribution plants, ships, chemical plants information systems designed for the control of energy systems automated control of water supply systems

5 Department of structural analysis Main tasks: Analysis of the effects and mechanisms of material degradation and evaluating the level of damage Irradiation damage monitoring of RPV materials using surveillance samples Monitoring corrosion and erosion of materials in selected systems of both primary and secondary systems Failure analysis of operational defects in equipment by evaluating material conditions

6 Description of SG WWER 440 The steam generator is horizontal shell-and-tube heat exchanger It consists of a pressure vessel, a horizontal heat exchange tube bundle, two vertical primary collectors, a feedwater piping system, moisture separators and steam collector Primary coolant enters the steam generator through a vertical collector, travels through the horizontal U-shaped submerged stainless steel tubing, and exits through a second vertical collector The tube ends penetrate the collector wall and are expanded using either a hydraulic or explosive expansion process and then welded at the collector inside wall surface Both the collectors and heat exchange tubes are made of Tistabilized austenitic stainless steel The steam generator vessel is a carbon steel horizontal cylinder consisting of forged shells, stamped elliptical ends and stamped branch pipes and hatches welded together

7 Main parameters of SG WWER 440 Parameter Unit Value Parameter Unit Value Thermal power [MW] Coolant flow rate in tubes [m/s] 2,71 Steam capacity [kg/s] 125 Specific heat flux (average) [kw/m 2 ] 89,23 Pressure of steam [MPa] 461 Total heat exchanging surface [m 2 ] 2576,6 Steam temperature [ C] 258,9 Total number of tubes 5536 Feedwater temperature [ C] Diameter of tube [mm] 16 Coolant temperature at steam generator inlet Coolant temperature at steam generator outlet [ C] 295 Thickness of tube walls [mm] 1,4 [ C] 267 Tube mean length [m] 9,26 Coolant flow rate [m 3 /h] 7100 Coolant pressure [MPa] 12,26 Pressure loss along the coolant path Steam humidity at steam generator outlet [MPa] 0,075 [%] 0,25

8 Scheme of SG WWER 440

9 Main welds of SG WWER 440

10 Main welds of SG WWER 440

11 Structural materials of SG WWER 440 Content of elements [mass %] Materials C Si Mn S P Cr Ni Cu Ti V Mo As Nb 08Ch18N10T 0,08 0,8 2,0 0,020 0, xC-0, K 1 0,19-0,26 0,2-0,4 0,75-1,0 0,030 0,030 0,3 0,3 0, K 0,16-0,24 0,15-0,30 0,35-0,65 0,040 0,040 0,25 0,25 0, ,17-0,24 0,17-0,37 0,35-0,65 0,040 0,040 0,3 0,3 0, Ch1MF 0,22-0,29 0,17-0,37 0,4-0,7 0,025 0,030 1,5-1, ,15-0,3O 0,25-0, Ch 3 0,36-0,44 0,17-0,37 0,5-0,8 0,025 0,025 0,8-1,1 0,3 0, Ch2MFA 0,22-0,27 0,17-0,37 0,3-0,6 0,025 0,025 2,8-3,3 0,4 0,3-0,25-0,35 0,6-0,8 0,8 - ChN35VT 4 0,12 0,60 1,0-2,0 0,015 0, ,1-1, W=2,8-3,5 UONI13/25 5 EB123-JE 0,11 0,18-0,5 0,65-1,2 0,033 0, ,0 - - AN-42 *, Sv08GSMT 6 0,12 0,3-0,9 0,8-1,6 0,030 0,030 0, ,2-0,4 - - EA 400/10T 5 0,1 0,6 1,15-3,1 0,025 0,025 16, ,3-0,75 2,0-3,5 - - EA 395/9 5 0,12 0,7 1,0-2,2 0,025 0,025 13, ,5-7,5 - - ZIO-8 5 0,12 1,0 2,7 0,030 0,030 23, , EA 898/21B 5 0,1 0,7 1,6-2,5 0,025 0,025 17, , ,3-1,05-0,8-1,05 Sv 04Ch19N11M3 6 0,1 1,0 0,8-2,0 0,020 0,030 15, ,5-3,0 - - Sv 07Ch25N13 7 0,09 1,2 0,8-2,0 0,020 0, , , Sv 08Ch19N10G2B 7 0,1 1,0 1,3-2,5 0,020 0,030 17,5-20 8, ,7-1,2 Sv 10Ch16N25M6 6 0,12 1,2 0,8-2,0 0,020 0,030 12, ,5 - -

12 Scheme of SG WWER steam generator body, 2 - primary cold leg collector, 3 - primary hot leg collector, 4 - manhole, 5 - heat exchanger tubes, 6 - vertical distance grid, 7- horizontal distance grid, 8 - feedwater pipeline, 9 - separator, 10 - perforated sheet, 11 - steam header, 12 - primary circuit header cover, 13 - secondary circuit header cover, 14 - cover seals for the primary and secondary circuit, 15 - secondary circuit seal cover monitoring location, 16 secondary circuit air vent, 17 - primary circuit seal cover monitoring location, 18 - primary circuit air vent, 19 - header periodic blowdown, 20 - steam generator periodic blowdown, 21 - steam generator permanent blowdown, 22 - nozzle, 23 - pipe unions for steam generator level checking

13 SGs in Slovak republic Bohunice V-1 NPP operation terminated 12 (2 x 6) (SU production) Bohunice V-2 NPP under operation 12 (2 x 6) (Czech rep. - Vitkovice production) Mochovce Unit 1,2 under operation 12 (2 x 6) (Czech rep. - Vitkovice production) Mochovce Unit 3,4 under construction 12 (2 x 6) (Czech rep. - Vitkovice production) Total 48

14 Operational problems and other issues a) Crack of the hot leg collector SG 32 in Bohunice NPP (1987) b) Cracking of ligaments on the hot leg collector SG 31 in Bohunice NPP (1987) c) Cracking of primary collector bolts d) Cracks on primary collector flange (1995) e) Leaky secondary circuit cover seal (2002) f) Leaky hot collector impulse lines (1999, 2007) g) Analysis of heat exchange tubes + brief summary of plugged tubes

15 ad c) Cracking of primary collector bolts Cracking of the primary collector bolts long-lasting problem Material austenitic stainless steel, type ChN35VT-VD with a high content of Ni Mechanical properties increased - result of complicated thermal treatment thermal hardening Example of the thermal treatment: > anealing 1000 C/50min/water > 1. hardening 840 C/10h/water > C/49h/water Unwanted result high level of sensitization to IG corrosion and to IGSCC Solution lower pre-stressing of the bolts replacement of the seal Ni ring Camprofile gaskets (grooved)

16 ad c) Cracking of primary collector bolts Material characterization: Chemical composition [mass %] C Mn Si P S Cr Ni Ti W ChN35VT-VD max. 0,12 1,0 až 2,0 max. 0,60 max. 0,025 max. 0,015 14,0 až 16,0 34,00 až 36,00 1,10 až 1,50 2,80 až 3,50 Mechanical properties t = 20 C t = 350 C Rp 0,2 Rm A 5 Z KCU2 Hardness Rp 0,2 [MPa] [MPa] [%] [%] [%] [HB] [Mpa] ChN35VT-VD min

17 ad c) Cracking of primary collector bolts Primary collector bolt - M48

18 ad c) Cracking of primary collector bolts Indication in nut part History: 5x / 10y 1y Cracks in thread heel Max. depth 8,0 mm

19 ad d) Cracks on primary collector flange Indications on primary collector flange and on first threads Material austenitic stainless steel type 08Ch18N10T Mechanism IG SCC Solution modification of the flange, precise tightening of the bolts

20 ad f) Leaky hot collector impulse lines

21 ad g) Analysis of heat exchange tubes Three tubes only were cut off and dragged out of SGs Only in one case the indication was found confirmed by metallographic evaluation Indication = repaired tube weld metal δ-ferrite magnetic phase All tubes were in a good condition, no ageing processes observed

22 ad g) Summary of plugged tubes Number of plugged tubes: Bohunice Unit #3 Bohunice Unit #4 SG No. Tubes Plugged Plugged [%] SG No. Tubes Plugged Plugged [%] SG ,192 % SG ,235 % SG ,090 % SG ,126 % SG ,885 % SG ,445 % Total Σ ,662 % SG ,054% SG ,000% SG ,018% SG ,199% SG ,000% SG ,018% Total Σ ,048% Mochovce Unit #1,2 less then Bohunice Unit #4

23 Technical upgrades of SG WWER New feedwater distribution system

24 Technical upgrades of SG WWER Primary collector covers with end stop

25 Technical upgrades of SG WWER Primary collector flange sealing Camprofile gasket (grooved) instead of Ni ring

26 Corrosion monitoring system in SG A simple equipment enables a long-term exposition of various samples inside the steam generator above the primary collector flange The samples are placed on a special holder into secondary circuit conditions The program has been interupted due to construction changes of primary collector cover XJ8 XJ9 XJ10 XJ11 XJ12 XJ13 XJ14 XU14 XU13 XU12 XU11 XU10 XU8 XU8 XJ1 XJ2 XJ3 XJ4 XJ5 XJ6 XJ7 XU7 XU6 XU5 XU4 XU3 XU2 XU1

27 Ageing management programs The ageing management program for steam generators, is supported by chemistry measurements to support evaluation of the impact of ageing on involved materials and material conditions, and reduce the intensity of the erosion - corrosion mechanism Cumulative fatigue damage is calculated and evaluated on a basis of temperature and pressure changes measured during operation: SG wall steam collector nozzle feedwater nozzle primary collector - hot leg

28 Fatigue damage of SG wall

29 Fatigue damage of steam collector nozzle

30 Fatigue damage of feedwater nozzle

31 Fatigue damage of hot leg primary collector

32 Conclusions SGs are very reliable components of NPPs type WWER 440 Total number of plugged tubes is still negligible Operational regime has been set to optimal parameters Used structural materials are stable in given standard conditions and make possible to extend the life of NPPs Applied or prepared ageing management programs can ensure safety and reliable operation for the planned and extended lifetime, too

CORROSION MONITORING SYSTEM IN THE SLOVAK REPUBLIC NUCLEAR POWER PLANTS

CORROSION MONITORING SYSTEM IN THE SLOVAK REPUBLIC NUCLEAR POWER PLANTS CORROSION MONITORING SYSTEM IN THE SLOVAK REPUBLIC NUCLEAR POWER PLANTS M. Brezina, L. Kupca VUJE Inc. Okruzna 5, 918 64 Trnava, Slovak Republic Email address of main author: brezina@vuje.sk Abstract.

More information

Application of Small Punch Testing Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP

Application of Small Punch Testing Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP Application of Small Punch Testing Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP Jana Petzová 1, a *, Martin Březina 1, b, Miloš Baľák 1, c, Ľudovít Kupča 1, e 1 VUJE, a.s.,

More information

Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors)

Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors) Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors) 1.3.2016 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at

More information

NUCLEAR POWER DELIVERIES. Ing. Martin Pecina, MBA General director VÍTKOVICE POWER ENGINEERING a. s ATOMEX Prague

NUCLEAR POWER DELIVERIES. Ing. Martin Pecina, MBA General director VÍTKOVICE POWER ENGINEERING a. s ATOMEX Prague Ing. Martin Pecina, MBA General director VÍTKOVICE POWER ENGINEERING a. s. 25. 26. 10. 2011 ATOMEX Prague We are the most significant Central European machinery group with strong position in selected segments

More information

VVER-440/213 - The reactor core

VVER-440/213 - The reactor core VVER-440/213 - The reactor core The fuel of the reactor is uranium dioxide (UO2), which is compacted to cylindrical pellets of about 9 height and 7.6 mm diameter. In the centreline of the pellets there

More information

Heat exchanger equipment of TPPs & NPPs

Heat exchanger equipment of TPPs & NPPs Heat exchanger equipment of TPPs & NPPs Lecturer: Professor Alexander Korotkikh Department of Atomic and Thermal Power Plants TPPs Thermal power plants NPPs Nuclear power plants Content Steam Generator

More information

5th Pan American Conference for NDT 2-6 October 2011, Cancun, Mexico. Systems for inspection and repair of WWER type steam generators

5th Pan American Conference for NDT 2-6 October 2011, Cancun, Mexico. Systems for inspection and repair of WWER type steam generators 5th Pan American Conference for NDT 2-6 October 2011, Cancun, Mexico Systems for inspection and repair of WWER type steam generators Adrian KOVALYK, Pavol JABLONICKY, Peter PILAT Division for diagnostics

More information

RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING

RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Science and Technology Journal of BgNS, Vol. 8, 1, September 2003, ISSN 1310-8727 RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Pavlin P. Groudev, Rositsa V. Gencheva,

More information

Eddy Current Inspection of WWER Steam Generator Tubes - Sensitivity of Bobbin Probe Technique

Eddy Current Inspection of WWER Steam Generator Tubes - Sensitivity of Bobbin Probe Technique ECNDT 2006 - Th.3.1.4 Eddy Current Inspection of WWER Steam Generator Tubes - Sensitivity of Bobbin Probe Technique Roman KRAJČOVIČ, Josef PLÁŠEK VUJE, a.s. Trnava, Slovak Republic Abstract. Eddy current

More information

Lifetime-Management and Operational Lifetime Extension at Paks Nuclear Power Plant

Lifetime-Management and Operational Lifetime Extension at Paks Nuclear Power Plant Transactions of the 17 th International Conference on Structural Mechanics in Reactor Technology (SMiRT 17) Prague, Czech Republic, August 17 22, 2003 Paper # D02-2 Lifetime-Management and Operational

More information

Systems for inspection and repair of WWER type steam generators. Adrian KOVALYK, Pavol JABLONICKY, Peter PILAT

Systems for inspection and repair of WWER type steam generators. Adrian KOVALYK, Pavol JABLONICKY, Peter PILAT Systems for inspection and repair of WWER type steam generators Adrian KOVALYK, Pavol JABLONICKY, Peter PILAT 2011 What we do... Reconstruction & modernization Nuclear safety Diagnostics & In-service inspection

More information

IAEA-TECDOC-1361 Assessment and management of ageing of major nuclear power plant components important to safety

IAEA-TECDOC-1361 Assessment and management of ageing of major nuclear power plant components important to safety IAEA-TECDOC-1361 Assessment and management of ageing of major nuclear power plant components important to safety Primary piping in PWRs July 2003 The originating Section of this publication in the IAEA

More information

Karel Böhm 1, Milan Brumovsky 2, Jirí Zd árek 2 ABSTRACT

Karel Böhm 1, Milan Brumovsky 2, Jirí Zd árek 2 ABSTRACT Karel BOHM, SUBJ (Czech Republic) Milan BRUMOVSKY, Jirí ZD ÁREK, Nuclear Research Institute (Czech Republic) Integrated surveillance specimen programme for WWER-1000/V-320 reactor pressure vessels Karel

More information

Overview, Irradiation Test and Mechanical Property Test

Overview, Irradiation Test and Mechanical Property Test IAE R&D Program Progress Report Development Project of Supercritical-water Cooled Power Reactors Overview, Irradiation Test and Mechanical Property Test Shigeki Kasahara Hitachi, Ltd. Toshiba Corp. Hokkaido

More information

Supercritical Water Reactor Review Meeting. Materials Issues

Supercritical Water Reactor Review Meeting. Materials Issues Supercritical Water Reactor Review Meeting Materials Issues Bill Corwin, Louis Mansur, Randy Nanstad, Arthur Rowcliffe, Bob Swindeman, Peter Tortorelli, Dane Wilson, Ian Wright Oak Ridge National Laboratory

More information

Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: PWR Vessel Internals

Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: PWR Vessel Internals IAEA-TECDOC-1557 Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: PWR Vessel Internals 2007 Update June 2007 IAEA SAFETY RELATED PUBLICATIONS IAEA SAFETY

More information

Qualification of NDT Systems in Hungary

Qualification of NDT Systems in Hungary ECNDT 2006 - We.1.4.2 of NDT Systems in Hungary George SOMOGYI, Hungarian Inspection Body, Budapest, Hungary Denes SZABO, Paks Nuclear Power Plant, Paks, Hungary Peter TRAMPUS, Trampus Consulting & Engineering,

More information

CONTENTS. Foreword... xiii Statements of Policy... xv Personnel... xvii Organization of Section III... xxvii Summary of Changes...

CONTENTS. Foreword... xiii Statements of Policy... xv Personnel... xvii Organization of Section III... xxvii Summary of Changes... CONTENTS Foreword... xiii Statements of Policy... xv Personnel... xvii Organization of Section III... xxvii Summary of Changes... xxxi Article NB-1000 Introduction... 1 NB-1100 Scope... 1 NB-1110 Aspects

More information

Relationship Between Design and Materials for Thermal Power Plants. S. C. Chetal

Relationship Between Design and Materials for Thermal Power Plants. S. C. Chetal Relationship Between Design and Materials for Thermal Power Plants S. C. Chetal Contents *Introduction to design codes *Operational life vs design basis life *Material selection basics *Materials for boiler

More information

Grades : SS 304, 304L, 304H, 316, 316L, 316LN, 316Ti, 309, 310S, 317L, 321, 347,

Grades : SS 304, 304L, 304H, 316, 316L, 316LN, 316Ti, 309, 310S, 317L, 321, 347, Unifit MetalloysInc (UMI) is one of the leading Manufacturer, Exporter and Supplier of High Quality Stainless Steel Pipes & Tubes, Flanges, Pipe Fittings, Round Bars, Sheets and Plates Etc.Stainless steel

More information

PSA ANALYSIS FOCUSED ON MOCHOVCE NPP SAFETY MEASURES EVALUATION FROM OPERATIONAL SAFETY POINT OF VIEW

PSA ANALYSIS FOCUSED ON MOCHOVCE NPP SAFETY MEASURES EVALUATION FROM OPERATIONAL SAFETY POINT OF VIEW International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: http://www.drustvo-js.si/port2001/ e-mail: PORT2001@ijs.si tel.:+ 386 1 588

More information

RPV DESIGN. FABRICATION AND MATERIALS

RPV DESIGN. FABRICATION AND MATERIALS RPV DESIGN. FABRICATION AND MATERIALS Marta Serrano. CIEMAT IAEA Training Workshop on Assessment of Degradation Mechanisms of Primary Components in Water Cooled Nuclear Reactors: Current Issues and Future

More information

TERMO-HYDRAULICS AND TERMO-MECHANICAL LOADING OF VVER-440 REACTOR PRESSURE VESSEL

TERMO-HYDRAULICS AND TERMO-MECHANICAL LOADING OF VVER-440 REACTOR PRESSURE VESSEL TERMO-HYDRAULICS AND TERMO-MECHANICAL LOADING OF VVER-440 REACTOR PRESSURE VESSEL G. Gálik 1, V. Kutiš 2, J. Paulech 3, V. Goga 4 Abstract: This article describes a pressure thermal shock simulation methodology

More information

Available online at ScienceDirect. Procedia Engineering 86 (2014 )

Available online at  ScienceDirect. Procedia Engineering 86 (2014 ) Available online at www.sciencedirect.com ScienceDirect Procedia Engineering 86 (2014 ) 308 314 1st International Conference on Structural Integrity, ICONS-2014 Guidelines for Integrity and Lifetime Assessment

More information

IAEA CONTRIBUTION TO ASSESSMENT AND MANAGEMENT OF STEAM GENERATOR AGEING

IAEA CONTRIBUTION TO ASSESSMENT AND MANAGEMENT OF STEAM GENERATOR AGEING IAEA CONTRIBUTION TO ASSESSMENT AND MANAGEMENT OF STEAM GENERATOR AGEING Introduction L. Kupca, K. S. Kang, International Atomic Energy Agency, Wagramer Strasse 5, PO Box 100, 1400 Wien, Austria Ageing

More information

Lifetime analysis of WWER Reactor Pressure Vessel Internals concerning material degradation

Lifetime analysis of WWER Reactor Pressure Vessel Internals concerning material degradation 20th International Conference on Structural Mechanics in Reactor Technology (SMiRT 20) Espoo, Finland, August 9-14, 2009 SMiRT 20-Division 2, Paper 1893 Lifetime analysis of WWER Reactor Pressure Vessel

More information

RADIATION DAMAGE IN HIGH Ni-WELD OF Ni-Cr-Mo-V TYPE

RADIATION DAMAGE IN HIGH Ni-WELD OF Ni-Cr-Mo-V TYPE RADIATION DAMAGE IN HIGH Ni-WELD OF Ni-Cr-Mo-V TYPE M.Brumovsky, M.Kytka, R.Kopriva UJV REZ, CZECH REPUBLIC Degradation of primary components of Pressurised Water Cooled Nuclear Reactors: IAEATechnical

More information

WWER-1000 STEAM GENERATOR INTEGRITY

WWER-1000 STEAM GENERATOR INTEGRITY IAEA-EBP-WWER-07 XA9949240 WWER-1000 STEAM GENERATOR INTEGRITY A PUBLICATION OF THE EXTRARUDGETARY PROGRAMME ON THE SAFETY OF WWER AND RRMK NUCLEAR POWER PLANTS July 1997 30-10 m INTERNATIONAL ATOMIC ENERGY

More information

EFFECT OF COLD DEFORMATION ON THE PROPERTIES OF NEW AUSTENITIC STAINLESS STEEL FOR BOILER SUPERHEATER TUBES

EFFECT OF COLD DEFORMATION ON THE PROPERTIES OF NEW AUSTENITIC STAINLESS STEEL FOR BOILER SUPERHEATER TUBES EFFECT OF COLD DEFORMATION ON THE PROPERTIES OF NEW AUSTENITIC STAINLESS STEEL FOR BOILER SUPERHEATER TUBES Šárka HERMANOVÁ a, Lenka DOBROVODSKÁ a, Ladislav KANDER b a VÍTKOVICE POWER ENGINEERING a.s.,

More information

Datasheet for Steel Grades Superalloys Incoloy 800

Datasheet for Steel Grades Superalloys Incoloy 800 This page is mainly introduced the Incoloy 800 Datasheet, including chemical information,mechanical properties, physical properties, mechanical properties, heat treatment, and Micro structure, etc. It

More information

View from the Penthouse

View from the Penthouse View from the Penthouse The DNFM Technical News Letter David N. French Metallurgists Ph: 502-955-9847 Fax: 502-957-5441 Web: www.davidnfrench.com Load Cycling As many coal-fired power plants designed for

More information

Examples Page 43 of 45

Examples Page 43 of 45 UHDE-STANDARD Welding UN WELDED JOINTS V416-01 for vessels and equipment Part 2 (M) Examples Page 43 of 45 Item Figure Application Requirements Notes Small pad type flanges of austenitic steel, clad vessel

More information

Joint ICTP/IAEA Workshop on Irradiation-induced Embrittlement of Pressure Vessel Steels November 2009

Joint ICTP/IAEA Workshop on Irradiation-induced Embrittlement of Pressure Vessel Steels November 2009 2067-11 Joint ICTP/IAEA Workshop on Irradiation-induced Embrittlement of Pressure Vessel Steels 23-27 November 2009 RPV design, manufacturing and materials Milan Brumovsky Nuclear Research Institute Rez

More information

SEMPELL HP PREHEATER PROTECTION VALVES TYPE AVS 4/5

SEMPELL HP PREHEATER PROTECTION VALVES TYPE AVS 4/5 Typical feed water heater isolation system FEATURES Body of forged steel Body in form piece to reduce a number of welds and fittings Body also available as a single block type thereby eliminating the nozzle

More information

Structural Integrity and NDE Reliability II

Structural Integrity and NDE Reliability II Structural Integrity and NDE Reliability II NDE Experience and Lessons Learnt from Recent EU Projects Focused on Assistance to the Armenian NPP L. Horacek, Nuclear Research Institute, Czech Republic ABSTRACT

More information

Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems

Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems Stress Corrosion Cracking Susceptibility of Austenitic Stainless Steels in Supercritical Water Conditions R. Novotny 1), P. Hähner 1), J. Siegl 2), S. Ripplinger 1), Sami Penttilä 3), Aki Toivonen 3) 1)

More information

China Yangzhou plant

China Yangzhou plant China Yangzhou plant TM Marcegaglia China 714,000 sqm total area 153,000 sqm first phase covered area EN norms production quality standards Precision tubes from carbon and stainless steel Beijing Yangzhou

More information

Submerged Arc Welding Consumables For Mild Steel and 490N/mm 2 Class High Tensile Strength Steel

Submerged Arc Welding Consumables For Mild Steel and 490N/mm 2 Class High Tensile Strength Steel 03 For Mild Steel and 490 Class High Tensile Strength Steel CA-502 UC-36 For single and multi-layer welding of mild and 490 class high tensile strength steel AWS F7A0-EH14 F7P0-EH14 For Mild Steel and

More information

IAEA-TECDOC Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators

IAEA-TECDOC Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators IAEA-TECDOC-1668 Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators Assessment and Management of Ageing of Major Nuclear Power Plant Components

More information

SCC of SG tubing and stainless steel (SS) pipes and welds (PWRs) - 1

SCC of SG tubing and stainless steel (SS) pipes and welds (PWRs) - 1 International Conference on Water Chemistry of Nuclear Reactor Systems, 11-14 October 2004, San Francisco, EPRI in co-operation with the IAEA Participation: more than 200 experts from 23 countries; IAEA

More information

GENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS

GENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS - June 2013 Addendum GENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS SUBSECTION "A" : GENERAL RULES SUBSECTION "B" : CLASS 1 COMPONENTS SUBSECTION "C" : CLASS 2 COMPONENTS SUBSECTION "D" : CLASS

More information

Boiling Water Reactor Vessel and Internals

Boiling Water Reactor Vessel and Internals Boiling Water Reactor Vessel and Internals Program Description Program Overview As boiling water reactors have aged, various forms of operation-limiting stress corrosion cracking have appeared, first in

More information

The Mutual Influence of Materials and Thermal-hydraulics on Design of SCWR Review of Results of the Project HPLWR Phase 2

The Mutual Influence of Materials and Thermal-hydraulics on Design of SCWR Review of Results of the Project HPLWR Phase 2 Institute of Nuclear Technology and Energy Systems The Mutual Influence of Materials and Thermal-hydraulics on Design of SCWR Review of Results of the Project HPLWR Phase 2 J. Starflinger, T. Schulenberg

More information

ASTM Standards B127 B160 B161 B162 B163 B164 B165 B166 B167 B168 B333 B335 B407 B408 B409 B423

ASTM Standards B127 B160 B161 B162 B163 B164 B165 B166 B167 B168 B333 B335 B407 B408 B409 B423 ASTM Standards : B127 Specification for Nickel-Copper Alloy (UNS N04400) Plate, Sheet, B160 Specification for Nickel Rod and Bar B161 Specification for Nickel Seamless Pipe and Tube B162 Specification

More information

Preliminary application of the draft code case for alloy 617 for a high temperature component

Preliminary application of the draft code case for alloy 617 for a high temperature component Journal of Mechanical Science and Technology Journal of Mechanical Science and Technology 22 (2008) 856~863 www.springerlink.com/content/1738-494x Preliminary application of the draft code case for alloy

More information

Status Update 18MW ILC Beam Dump Design

Status Update 18MW ILC Beam Dump Design Status Update 18MW ILC Beam Dump Design John Amann Mechanical Engineer SLAC National Accelerator Laboratory LC Beam Delivery Systems Global Design Effort 1 Outline Introduction. Mechanical Design Concept.

More information

EFFECT OF MATERIALS HETEROGENEITIES ON MICROSTRUCTURE AND MECHANICAL PROPERTIES AT IRRADIATED STATE

EFFECT OF MATERIALS HETEROGENEITIES ON MICROSTRUCTURE AND MECHANICAL PROPERTIES AT IRRADIATED STATE Training School. 3-7 September 2018 Polytechnic University of Valencia (Spain) EFFECT OF MATERIALS HETEROGENEITIES ON MICROSTRUCTURE AND MECHANICAL PROPERTIES AT IRRADIATED STATE Hans-Werner Viehrig This

More information

September 2012 EGO Stainless Steel Tubes - Technical Details page 1 of 9

September 2012 EGO Stainless Steel Tubes - Technical Details page 1 of 9 Table of contents 1) Branches/applications for EGO Stainless Steel Tubes 2) Standard tube program 3) Materials 4) Chemical analysis of tube materials 5) Technology 6) Quality information page 1 of 9 1)

More information

Technical University of Sofia, Department of Thermal and Nuclear Power Engineering, 8 Kliment Ohridski Blvd., 1000 Sofia, Bulgaria

Technical University of Sofia, Department of Thermal and Nuclear Power Engineering, 8 Kliment Ohridski Blvd., 1000 Sofia, Bulgaria BgNS TRANSACTIONS volume 20 number 2 (2015) pp. 143 149 Comparative Analysis of Nodalization Effects and Their Influence on the Results of ATHLET Calculations of VVER-1000 Coolant Transient Benchmark Phase

More information

IAEA-TECDOC Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

IAEA-TECDOC Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals IAEA-TECDOC-1471 Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals October 2005 IAEA SAFETY RELATED PUBLICATIONS IAEA SAFETY

More information

Phenomenon of (irradiation assisted) stress corrosion cracking for internals of PWR & BWR systems

Phenomenon of (irradiation assisted) stress corrosion cracking for internals of PWR & BWR systems Phenomenon of (irradiation assisted) stress corrosion cracking for internals of PWR & BWR systems Rdk Radek Novotny & Luigi Lii Db Debarberisb Institute for Energy (IE) Petten, The Netherlands http://www.jrc.ec.europa.eu

More information

SPECIFIC DEGRADATIONS OF VVER-1000

SPECIFIC DEGRADATIONS OF VVER-1000 SPECIFIC DEGRADATIONS OF VVER-1000 (in view of lifetime extension) Dimitar Popov Kozloduy NPP, Bulgaria IAEA Technical Meeting on Degradation of Primary Components of PW cooled NPPs, Vienna, 05-08 Nov,

More information

Profile SFR-52 SWAT JAPAN. Japan Atomic Energy Agency, 4002 Narita, Oarai-machi, Ibaraki, Japan.

Profile SFR-52 SWAT JAPAN. Japan Atomic Energy Agency, 4002 Narita, Oarai-machi, Ibaraki, Japan. Profile SFR-52 SWAT JAPAN GENERAL INFORMATION NAME OF THE FACILITY ACRONYM COOLANT(S) OF THE FACILITY LOCATION (address): OPERATOR CONTACT PERSON (name, address, institute, function, telephone, email):

More information

TMK HYDRA ART INSIDE SKIVED AND ROLLER BURNISHED COLD FINISHED SEAMLESS STEEL TUBES and PISTON RODS TUBES FOR HYDRAULIC CYLINDERS

TMK HYDRA ART INSIDE SKIVED AND ROLLER BURNISHED COLD FINISHED SEAMLESS STEEL TUBES and PISTON RODS TUBES FOR HYDRAULIC CYLINDERS TMK HYDRA ART INSIDE SKIVED AND ROLLER BURNISHED COLD FINISHED SEAMLESS STEEL TUBES and PISTON RODS TUBES FOR HYDRAULIC CYLINDERS TMK is one of the world s leading producers of tubular products for the

More information

Section 4. Annex to IAEA Request for Proposal STATEMENT OF WORK. Primary/Secondary cooling water Heat Exchangers

Section 4. Annex to IAEA Request for Proposal STATEMENT OF WORK. Primary/Secondary cooling water Heat Exchangers Section 4 Annex to IAEA Request for Proposal 25919 STATEMENT OF WORK Primary/Secondary cooling water Heat Exchangers WWR-SM Research Reactor, UZBEKISTAN Page 1 of 17 1. Scope 1.1 This Statement of Work

More information

Supporting Deterministic T-H Analyses for Level 1 PSA

Supporting Deterministic T-H Analyses for Level 1 PSA Supporting Deterministic T-H Analyses for Level 1 PSA ABSTRACT SLAVOMÍR BEBJAK VUJE, a.s. Okružná 5 918 64 Trnava, Slovakia slavomir.bebjak@vuje.sk TOMÁŠ KLIMENT VUJE, a.s. Okružná 5 918 64 Trnava, Slovakia

More information

Analysis of water production costs of a nuclear desalination plant with a nuclear heating reactor coupled with MED processes

Analysis of water production costs of a nuclear desalination plant with a nuclear heating reactor coupled with MED processes Desalination 190 (2006) 287 294 Analysis of water production costs of a nuclear desalination plant with a nuclear heating reactor coupled with MED processes Shaorong Wu Institute of Nuclear Energy Technology,

More information

We are IntechOpen, the world s leading publisher of Open Access books Built by scientists, for scientists. International authors and editors

We are IntechOpen, the world s leading publisher of Open Access books Built by scientists, for scientists. International authors and editors We are IntechOpen, the world s leading publisher of Open Access books Built by scientists, for scientists 4,100 116,000 120M Open access books available International authors and editors Downloads Our

More information

SANDVIK SPRINGFLEX SF SPRING WIRE WIRE

SANDVIK SPRINGFLEX SF SPRING WIRE WIRE SANDVIK SPRINGFLEX SF SPRING WIRE WIRE DATASHEET Sandvik Springflex SF is a duplex (austenitic-ferritic) stainless steel especially designed for spring applications with extremely high fatigue requirements.

More information

SANDVIK SAF 2304 TUBE AND PIPE, SEAMLESS

SANDVIK SAF 2304 TUBE AND PIPE, SEAMLESS SANDVIK SAF 2304 TUBE AND PIPE, SEAMLESS DATASHEET Sandvik SAF 2304 is a lean duplex (austenitic-ferritic) stainless steel characterized by the following properties: Very good resistance to stress corrosion

More information

Typical Feed Water Heater Isolation System.

Typical Feed Water Heater Isolation System. Typical Feed Water Heater Isolation System. Features and Benefits In the event of a defective tube system these valves are designed to stop the feed of high pressure feedwater and to bypass it around a

More information

Latest SCC Issues of Core Shroud and Recirculation Piping in Japanese BWRs

Latest SCC Issues of Core Shroud and Recirculation Piping in Japanese BWRs Transactions of the 7 th International Conference on Structural Mechanics in Reactor Technology (SMiRT 7) Prague, Czech Republic, August 7, 3 Paper # WG- Latest SCC Issues of Core Shroud and Recirculation

More information

Structural Performance of next-generation nuclear components: lessons from the UK's R5 and R6 structural integrity assessment procedures

Structural Performance of next-generation nuclear components: lessons from the UK's R5 and R6 structural integrity assessment procedures Structural Performance of next-generation nuclear components: lessons from the UK's R5 and R6 structural integrity assessment procedures MATISSE Workshop on cross-cutting issues in structural materials

More information

K-FLEX SOLAR SYSTEM SOLAR SYSTEM

K-FLEX SOLAR SYSTEM SOLAR SYSTEM K-FLEX SOLAR SYSTEM SOLAR SYSTEM K-FLEX offers a line of products designed specifically for solar heating, renewable energy sources used to produce hot water using solar energy. The product range consists

More information

Structural Integrity and NDE Reliability I

Structural Integrity and NDE Reliability I Structural Integrity and NDE Reliability I Assessment of Failure Occurrence Probability as an Input for RI-ISI at Paks NPP R. Fótos, University of Miskolc, Hungary L. Tóth, P. Trampus, University of Debrecen,

More information

Stock program Stainless steel bars Stainless, acid- and heat-resistant steel

Stock program Stainless steel bars Stainless, acid- and heat-resistant steel Stock program Stainless steel bars Stainless, acid- and heat-resistant steel 2 CONTENT Content Steel grades EN Short name AISI UNS Microstructure Page 1.4021 X20Cr13 420 S42000 M 6 / 7 1.4028 X30Cr13 420

More information

PTS re-evaluation project for Czech NPPs

PTS re-evaluation project for Czech NPPs PTS re-evaluation project for Czech NPPs Vladislav Pištora, Miroslav Žamboch, Pavel Král, Ladislav Vyskočil Fourth International Conference on Nuclear Power Plant Life Management 23 27 October 2017 Lyon,

More information

Stainless Steel & Stainless Steel Fasteners Chemical, Physical and Mechanical Properties

Stainless Steel & Stainless Steel Fasteners Chemical, Physical and Mechanical Properties Stainless Steel & Stainless Steel Fasteners Chemical, Physical and Mechanical Properties Stainless steel describes a family of steels highly resistant to tarnishing and rusting that contain at least two

More information

BRODER METALS GROUP HTS 316

BRODER METALS GROUP HTS 316 Broder Metals supply 316 and 316L stainless cold drawn in B8M Class 2, B8M Class 2B, A4-70 and A4-80. A4-70 and A4-80 are named by their minimum tensile strength to be met by the material (70 & 80) by

More information

IAEA-TECDOC Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals

IAEA-TECDOC Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals IAEA-TECDOC-1119 Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals October 1999 The originating Section of this publication in the IAEA

More information

Takeyuki INAGAKI, Cesilla TOTH, Radim HAVEL, IAEA Nuclear Safety Abstract

Takeyuki INAGAKI, Cesilla TOTH, Radim HAVEL, IAEA Nuclear Safety Abstract Takeyuki INAGAKI, Cesilla TOTH, Radim HAVEL, IAEA Nuclear Safety IAEA guidance documents on Ageing Management of key safety components in nuclear power plants and Safety Knowledge-base on ageing and long-term

More information

30ChGSA Included in 13 standards (CIS Countries)

30ChGSA Included in 13 standards (CIS Countries) Standards GOST 10702-78 GOST 11268-76 GOST 12132-66 GOST 21729-76 GOST 23270-89 GOST 4543-71 GOST 8731-74 GOST 8733-87 GOST R 54159-10 TU 14-1-1213-75 TU 14-1-1409-75 TU 14-1-4118-76 TU 14-4-385-73 Steel

More information

SANDVIK NANOFLEX STRIP STEEL

SANDVIK NANOFLEX STRIP STEEL SANDVIK NANOFLEX STRIP STEEL DATASHEET Sandvik Nanoflex is a precipitation hardening, austenitic stainless steel specifically designed for applications requiring high strength and good ductility. Mechanical

More information

CANDU Safety #1 - CANDU Nuclear Power Plant Design Dr. V.G. Snell Director Safety & Licensing

CANDU Safety #1 - CANDU Nuclear Power Plant Design Dr. V.G. Snell Director Safety & Licensing CANDU Safety #1 - CANDU Nuclear Power Plant Design Dr. V.G. Snell Director Safety & Licensing 24/05/01 8:14 AM CANDU Safety - #1 - CANDU Design.ppt Rev. 1 vgs 1 What Accident is This? 28 killed, 36 injured,

More information

Research Article Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

Research Article Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor Science and Technology of Nuclear Installations, Article ID 18182, 6 pages http://dx.doi.org/1.1155/214/18182 Research Article Investigation of TASS/SMR Capability to Predict a Natural Circulation in the

More information

Guidelines for the Construction of Pressure Vessel Type Tanks Intended for the Transportation of Anhydrous Ammonia at Ambient Temperatures

Guidelines for the Construction of Pressure Vessel Type Tanks Intended for the Transportation of Anhydrous Ammonia at Ambient Temperatures (1992) (1992/Corr.) Guidelines for the Construction of Pressure Vessel Type Tanks Intended for the Transportation of Anhydrous Ammonia at Ambient Temperatures 1. Scope 1.1 These Guidelines complement the

More information

ENSA ENWESA SERVICES INDEX

ENSA ENWESA SERVICES INDEX ENSA ENWESA SERVICES INDEX 1. ENGINEERING, DESIGN AND TECHNOLOGY SERVICES (Ensa / ENWESA) Ensa has extensive experience designing nuclear components and pressure vessels since 1973 according to international

More information

Feedwater Flow Measurement with Venturi and Comparison to the other Parameters in NPP Krško

Feedwater Flow Measurement with Venturi and Comparison to the other Parameters in NPP Krško Feedwater Flow Measurement with Venturi and Comparison to the other Parameters in NPP Krško ABSTRACT Vinko Planinc, Aljoša Šumlaj, Robert Rostohar, Dejvi Kadivnik Nuclear Power Plant Krško Vrbina 12, SI-8270

More information

SAPAG. A full range of API 526 flanged safety relief valves for process applications, gas, steam and liquid. Safety Relief Valve Series 8100/8200

SAPAG. A full range of API 526 flanged safety relief valves for process applications, gas, steam and liquid. Safety Relief Valve Series 8100/8200 SAPAG A full range of API 526 flanged safety relief valves for process applications, gas, steam and liquid. Features Full compliance to API RP 520 and standards 526, 527. Certified ASME Section VIII on

More information

The Effect of Dissolved Oxygen on Stress Corrosion Cracking of 310S in SCW

The Effect of Dissolved Oxygen on Stress Corrosion Cracking of 310S in SCW CNNC NPIC The Effect of Dissolved Oxygen on Stress Corrosion Cracking of 310S in SCW Liu Jinhua Bin Gong Outline 1 Introduction 2 Experimental 3 Results and Discussion 4 Conclusions 5 Future Work 2016/10/28

More information

EU considerations on Design and Qualification of Plasma Facing Components for ITER

EU considerations on Design and Qualification of Plasma Facing Components for ITER EU considerations on Design and Qualification of Plasma Facing Components for ITER Patrick Lorenzetto, F4E Barcelona with inputs from B. Riccardi (F4E), V. Barabash and M. Merola (ITER IO) on Readiness

More information

Weld Distortion Control Methods and Applications of Weld Modeling

Weld Distortion Control Methods and Applications of Weld Modeling Weld Distortion Control Methods and Applications of Weld Modeling F. W. Brust Paul Scott ABSTRACT A virtual fabrication technology (VFT) modeling procedure is introduced in this paper. It is a state-of-theart

More information

OperatiOn and safety report Of MOchOvce and BOhunice v2 nuclear power plants

OperatiOn and safety report Of MOchOvce and BOhunice v2 nuclear power plants 2016 OperatiOn and safety report Of MOchOvce and BOhunice v2 nuclear power plants The company is certified according to three management systems: Certificate stn en iso 9001:2008 Quality management system

More information

BASIC OF PIPING MATERIAL

BASIC OF PIPING MATERIAL By: Miduk Aritonang 1. MATERIAL MATERIAL METAL NON METAL: - Plastic : Fibre Rainforced Plastic (FRP), etc - Polymer : PE, PVC, PTFE/Teflon, etc FERRO: - Carbon Steel (CS) - Low Temperature Carbon Steel

More information

Annex to the Accreditation Certificate D-PL according to DIN EN ISO/IEC 17025:2005

Annex to the Accreditation Certificate D-PL according to DIN EN ISO/IEC 17025:2005 Deutsche Akkreditierungsstelle GmbH Annex to the Accreditation Certificate D-PL-19221-01-00 according to DIN EN ISO/IEC 17025:2005 Period of validity: 11.04.2017 to 19.05.2019 Holder of certificate: DEKRA

More information

Materials Aging Management Programs at

Materials Aging Management Programs at Materials Aging Management Programs at Nuclear Power Plants in the United States Timothy J. Griesbach Structural Integrity Associates IAEA 2 nd International Symposium on Nuclear Power Plant Life Mgmt.

More information

Code. 1) Code. Standard or Specification. Bolting

Code. 1) Code. Standard or Specification. Bolting Code 1. Code 1) Code (1) ASME B 31.1 : Power Piping (2) ASME B 31.3 : Process Piping (3) ASME B 31.4 : Pipeline Transportation Systems for Liquid Hydrocarbons and Other Liquids 2) Material Code Bolting

More information

IAEA-TECDOC Reference manual on the IAEA JRQ correlation monitor steel for irradiation damage studies

IAEA-TECDOC Reference manual on the IAEA JRQ correlation monitor steel for irradiation damage studies IAEA-TECDOC-123 Reference manual on the IAEA JRQ correlation monitor steel for irradiation damage studies July 21 The originating Section of this publication in the IAEA was: Nuclear Power Engineering

More information

DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION

DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION VIENNA OKTOBER 3-6, 2016 1 ANPP * ANPP is located in the western part of Ararat valley 30 km west of Yerevan close to

More information

ASTM A420-Standard Specification for Piping Fittings of Wrought Carbon Steel and Alloy Steel for Low-Temperature Service

ASTM A420-Standard Specification for Piping Fittings of Wrought Carbon Steel and Alloy Steel for Low-Temperature Service ASTM A420-Standard Specification for Piping Fittings of Wrought Carbon Steel and Alloy Steel for Low-Temperature Service 1. Scope 1.1 This specification 2 covers wrought carbon steel and alloy steel fittings

More information

Laser applications in nuclear power plants

Laser applications in nuclear power plants PRAMANA c Indian Academy of Sciences Vol. 82, No. 1 journal of January 2014 physics pp. 135 141 Laser applications in nuclear power plants D N SANYAL Remote Tooling Section, Technology Development Group,

More information

GTAW Wires for Stainless Steel

GTAW Wires for Stainless Steel GTAW Wires for Stainless Steel Products Classification Typical chemical composition (wt %) AWS JIS EN C Mn Si Cr Ni Mo others KTS-307Si W 18 8 Mn 0.06 6.20 0.41 19.30 8.10 KTS-307HM W 18 8 Mn 0.06 6.20

More information

RESULTS OF THE GRADUAL UPGRADING AT BOHUNICE WWER - 440/230 NPP

RESULTS OF THE GRADUAL UPGRADING AT BOHUNICE WWER - 440/230 NPP RESULTS OF THE GRADUAL UPGRADING AT BOHUNICE WWER - 440/230 NPP P. Krupa Ingeneer, e-mail: Krupa_Peter@ebo.seas.sk Bohunice NPPs Introduction The centre of upgrading activities in VVER NPP is clearly in

More information

BODY OF KNOWLEDGE API-653 ABOVEGROUND STORAGE TANK INSPECTOR CERTIFICATION EXAMINATION

BODY OF KNOWLEDGE API-653 ABOVEGROUND STORAGE TANK INSPECTOR CERTIFICATION EXAMINATION BODY OF KNOWLEDGE API-653 ABOVEGROUND STORAGE TANK INSPECTOR CERTIFICATION EXAMINATION October 2009 (Replaces November 2008) API Authorized Aboveground Storage Tank Inspectors must have a broad knowledge

More information

Maintaining Nuclear Reactor Components Performance and Integrity for Plant Safety

Maintaining Nuclear Reactor Components Performance and Integrity for Plant Safety Boosting your Excellence through Knowledge and Training Maintaining Nuclear Reactor Components Performance and Integrity for Plant Safety Academy Academy Academy Seminar objective Academy Generally, well-run

More information

Cast steel: Group of ASTM standards for steel castings and forgings

Cast steel: Group of ASTM standards for steel castings and forgings Cast steel: Group of ASTM standards for steel castings and forgings Abstract: This group of ASTM specifications covers standard properties of steel and iron castings and forgings for valves, flanges, fittings,

More information

Visual Examination Assisted; RCA of Cracks In Thin Walled Duplex Stainless Steel Welds

Visual Examination Assisted; RCA of Cracks In Thin Walled Duplex Stainless Steel Welds More info about this article: http://www.ndt.net/?id=22345 Abstract: Visual Examination Assisted; RCA of Cracks In Thin Walled Duplex Stainless Steel Welds Kuldip KUMBEPHALKAR,P B PATIL, Bhagwan SHINDE

More information

Pressurized Water Reactor Materials Reliability Program (QA)

Pressurized Water Reactor Materials Reliability Program (QA) Pressurized Water Reactor Materials Reliability Program (QA) Program Description Program Overview Stress corrosion cracking and general environmental corrosion of reactor coolant system components have

More information

W-19 Steel Forgings for Continuous Grain-Flow Semi-Built Crankshafts

W-19 Steel Forgings for Continuous Grain-Flow Semi-Built Crankshafts Guideline No. W-19 (201705) W-19 Steel Forgings for Continuous Grain-Flow Semi-Built Crankshafts Issued date: May 9, 2017 China Classification Society Foreword This Guideline is a part of CCS Rules, which

More information

Stainless precision steel strip. With the best features.

Stainless precision steel strip. With the best features. Stainless precision steel strip. With the best features. Chemical Composition The currently deliverable rust, acid and heat-resistant stainless steels. Plant Abbreviated EN 188-2 EN 195 Material no. Comparable

More information