Light water reactor pressure vessel surveillance dosimetry improvement program. Volume 2. Postirradiation notch ductility and tensile strength determinations for PSF simulated surveillance and through-wall specimen capsules
The Program as irradiated mechanical property test specimens of several steels in a pressure vessel wall/thermal shield mock-up facility part of a broad NRC effort to develop key neutron physics-dosimetry-metallurgy correlations for making highly accurate projections of radiation-induced embrittlement to reactor vessels. The steels studied represent a wide range of radiation embrittlement sensitivities and include plates, forgings and submerged arc weld deposits (two each). This report presents notch ductility and tensile properties information developed with specimen irradiations at simulated surveillance and in-wall locations. The irradiations were conducted at 288/sup 0/C; neutron fluences were typical of vessel end-of-life conditions. Data comparisons are used to illustrate the toughness gradient produced by irradiation and to assess the relative irradiation effect at surveillance vs. throughwall positions. The postirradiation toughness gradient between vessel surface and midwall locations was small (31/sup 0/C or less) for five of the six materials. Tensile test observations support the notch ductility trend indications. Simulated surveillance capsule irradiations reproduced reasonably well the embrittlement at vessel inner surface and quarter wall thickness positions in almost all cases. The primary exceptiosn to both trends were provided by the steel having the highest embrittlement sensitivity (0.23% Cu, 1.58% Ni weld deposit). Aggregate results for this material suggest an independent contribution of high (> 1%) nickel contents to radiation sensitivity development and a weld susceptibility to long term time-at-temperature effects.
- Research Organization:
- Materials Engineering Associates, Inc., Lanham, MD (USA)
- OSTI ID:
- 7140384
- Report Number(s):
- NUREG/CR-3295-Vol.2; MEA-2017-Vol.2; ON: DE84901094
- Country of Publication:
- United States
- Language:
- English
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Light water reactor pressure vessel surveillance dosimetry improvement program. Volume 1. Notch ductility and fracture toughness degradation of A 302-B and A 533-B reference plates from PSF simulated surveillance and through-wall irradiation capsules
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Related Subjects
360106* -- Metals & Alloys-- Radiation Effects
ALLOYS
CARBON STEELS
CONTAINERS
EMBRITTLEMENT
IRON ALLOYS
IRON BASE ALLOYS
IRRADIATION
JOINTS
NEUTRON FLUENCE
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
RADIATION EFFECTS
REACTORS
STEEL-ASTM-A302
STEEL-ASTM-A533-B
STEELS
WATER COOLED REACTORS
WELDED JOINTS