Abstract

The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPPs) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterization. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I); irradiated and recovery annealed (IA); and irradiated, recovery annealed, and re-irradiated (IAI). The working program is focused on the characterization of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the American Society for Testing of Materials (ASTM) Test Standard E1921–08 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step, the trepans taken from the RPV Greifswald unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the master curve (MC) approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behavior. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on subsize Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy-V transition temperature TT41J estimated with results of subsize specimens after the recovery annealing was confirmed by the testing of standard Charpy-V-notch specimens. The evaluated TT41J shows a better accordance with the irradiation fluence along the wall thickness than the master curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during two campaign operations can be assumed to be low for the weld and base metal.

1.
Konheiser
,
J.
,
Rindelhardt
,
U.
,
Viehrig
,
H. -W.
,
Böhmer
,
B.
, and
Gleisberg
,
B.
, 2006, “
Pressure Vessel Investigations of the Former Greifswald NPP: Fluence Calculations and Nb Based Fluence Measurements
,”
ICONE14/FEDSM2006 Proceedings
, Paper No. ICONE 14-89578.
2.
Böhmer
,
B.
,
Böhmert
,
J.
,
Müller
,
G.
,
Rindelhardt
,
U.
, and
Utke
,
H.
, 1999, “
Embrittlement Studies of the Reactor Pressure Vessel of the Greifswald −440 Reactors
,” Technical Report No. NUCRUS96601.
3.
Brumovský
,
M.
,
Valo
,
M.
,
Kryukov
,
A.
,
Gillemot
,
F.
,
Debarberis
,
L.
, and
Kang
,
K.
, 2005, “
Guidelines for Prediction of Irradiation Embrittlement of Operating WWER-440 Reactor Pressure Vessels
,” IAEA-TECDOC-1442,
IAEA
,
Vienna
.
4.
ASTM
, 2006, “
ASTM E399: Standard Test Method for Plane-Strain Fracture Toughness of Metallic Materials
,”
Annual Book of ASTM Standards
,
ASTM
,
West Conshohocken, PA
.
5.
2003, DIN EN 10045-1 (1991): Metallic Materials: Charpy Impact Test; Part 1, DIN-Taschenbuch 19, Werkstoffprüfnormen für metallische Werkstoffe 1, Beut Verlag GmbH.
6.
ASTM
, 2008, “
ASTM E1921–08: Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range
,”
Annual Book of ASTM Standards
,
ASTM
,
West Conshohocken, PA
.
7.
Ahlstrand
,
R.
,
Klausnitzer
,
E. N.
,
Langer
,
D.
,
Leitz
,
Ch.
,
Pastor
,
D.
, and
Valo
,
M.
, 1993, “
ASTM STP 1170: Evaluation of the Recovery Annealing of the Reactor Pressure Vessel of NPP Nord (Greifswald) Units 1 and 2 by Means of Subsize Impact Specimens
,”
Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review
, Vol.
4
,
L. E.
Steel
, ed.,
American Society for Testing and Materials
,
Philadelphia
, pp.
321
343
.
8.
Viehrig
,
H. -W.
, and
Schuhknecht
,
J.
, 2009, “
Fracture Mechanics Characterisation of the WWER-440 Reactor Pressure Vessel Core Welding Seam
,”
Int. J. Pressure Vessels Piping
0308-0161,
86
, pp.
239
245
.
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