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Mager, T.R.
Westinghouse Electric Corp., Pittsburgh, Pa. (USA). PWR Systems Div1970
Westinghouse Electric Corp., Pittsburgh, Pa. (USA). PWR Systems Div1970
AbstractAbstract
No abstract available
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Source
Oct 1970; 39 p; For Oak Ridge National Lab., Tenn.
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AbstractAbstract
[en] ASME Code procedure for evaluating the acceptability of flaws detected during in-service inspection is revised. Critical crack size for instability is proposed as criteria for detected flaws in operating plants
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International Atomic Energy Agency, Vienna (Austria). International Working Group on Reliability of Reactor Pressure Components; p. 265-266; 1976; p. 265-266; Specialists' meeting on fracture mechanics applications: Implications of detected flaws; Winterthur, Switzerland; 3 - 5 Dec 1975
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Report
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Conference
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Mager, T.R.
Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Technology Div1982
Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Technology Div1982
AbstractAbstract
[en] Program materials were three weldments fabricated from A533 Grade B class 1 plate material and Mn Mo Ni weld wire. Specimens fabricated from the three submerged arc weldments included Type A Charpy V-notch impact, small size tensile, and 1/2T compact tension specimens. After encapsulation, the specimens were irradiated at the UVAR to two fluence levels, 8 x 1018 n/cm2 and 1.5 x 1019 n/cm2 (E > 1 MeV). Specimens were subjected to sequences of irradiation and anneals and then tested. Metallurgial/mechanistic analyses were also performed. It was concluded that excellent recovery of all properties could be achieved by annealing at greater than or equal to 8500F (4540C) for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the expected rate. Several drawbacks were identified for wet thermal annealing. A conceptual dry in-situ thermal annealing procedure was developed for thermal annealing embrittled reactor vessels
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Nov 1982; 499 p; Available from NTIS, PC E17/MF $10.25 as DE83900744
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Report
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Numerical Data
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Mager, T.R.
Westinghouse Electric Corp., Pittsburgh, Pa. (USA)1971
Westinghouse Electric Corp., Pittsburgh, Pa. (USA)1971
AbstractAbstract
No abstract available
Primary Subject
Source
17 Mar 1971; 10 p; 5. annual information meeting of the heavy section steel technology program; Oak Ridge, Tenn; 25 Mar 1971
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Report
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Conference
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Mager, T.R.
Westinghouse Electric Corp., Pittsburgh, Pa. (USA)1970
Westinghouse Electric Corp., Pittsburgh, Pa. (USA)1970
AbstractAbstract
No abstract available
Primary Subject
Source
1970; 12 p; 5. annual information meeting of the heavy section steel technology program; Oak Ridge, Tenn; 25 Mar 1971
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Report
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Conference
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AbstractAbstract
[en] Westinghouse has utilized precipitation hardening nickel base alloys for many years under reactor operating environment in a variety of applications. Inconel 600, 718 and X-750 has been used for coil springs, hold down springs, and in general all types of springs in the reactor internals. In addition, these alloys have been used as core support clevis material, all types of internal bolting and for guide tube support pins. Westinghouse used these nickel base precipitation hardening alloys for many years without an indication of stress corrosion cracking (SCC) phenomenon. It wasn't until 1978 that the first instance of SCC failure of Inconel X-750 occurred in a Westinghouse designed nuclear steam supply system. The failures of Inconel X-750 guide tube support pins prompted extensive material characterization and stress corrosion cracking test programs throughout the world. Westinghouse's testing program resulted in major changes in the Materials Specification for procurement of Inconel X-750. The changes included chemistry control, a well defined heat treatment, and closer control on grain size and micro-structure. This paper presents the evolution of material specifications as they pertain to inconel X-750. Test results that suggest that testing in steam at 400 degree C can be utilized to rank precipitation hardening alloys as to their susceptibility to SCC, and recommendations as to other alloy systems for reactor internals fasteners and bolting. 18 figs
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Gennaro, M.S.; Nelson, J.L.; Electric Power Research Inst., Palo Alto, CA (USA); Stone and Webster Engineering Corp., Boston, MA (USA); 578 p; May 1989; p. 8.1-8.18; Workshop on advanced high-strength materials; Clearwater Beach, FL (USA); 12-13 Mar 1986; CONF-8603261--; Research Reports Center, Box 50490, Palo Alto, CA 94303
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ALLOYS, ALUMINIUM ADDITIONS, CARBON ADDITIONS, CHEMICAL REACTIONS, CHROMIUM ALLOYS, CORROSION, CORROSION RESISTANT ALLOYS, CRYSTAL STRUCTURE, HEAT RESISTING ALLOYS, INCONEL ALLOYS, IRON ALLOYS, IRON BASE ALLOYS, MATERIALS, MOLYBDENUM ALLOYS, NICKEL ALLOYS, NICKEL BASE ALLOYS, NIOBIUM ADDITIONS, NIOBIUM ALLOYS, REACTORS, SIZE, STEELS, TITANIUM ADDITIONS, TITANIUM ALLOYS
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Mager, T.R.
Westinghouse Electric Corp., Pittsburgh, Pa. (USA). PWR Systems Div1970
Westinghouse Electric Corp., Pittsburgh, Pa. (USA). PWR Systems Div1970
AbstractAbstract
No abstract available
Primary Subject
Source
Nov 1970; 31 p; For Oak Ridge National Lab., Tenn.
Record Type
Report
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Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Mager, T.R.
Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Technology Div1983
Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Technology Div1983
AbstractAbstract
[en] An EPRI sponsored program was carried out by Westinghouse to determine the extent of fracture toughness recovery as a function of annealing time and temperature for neutron embrittlement sensitive reactor vessel material and to develop an optimal thermal anneal procedure for field applications. Program materials were three weldments fabricated by Combustion Engineering, Inc., from the same heat of A533 Grade B Class 1 plate material and the same heat of MnMoNi weld wire. The only variables were the target copper level and the welding flux which was Linde Grade 80 and Linde 0091. Weldments of 0.22, 0.36, and 0.41 wt % copper were produced. It was concluded from this study that excellent recovery of all properties could be achieved by annealing at 8500F (4540C) and above for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the rate which would have been expected had no anneal been performed. System limitations were identified for both wet and dry annealing methods
Original Title
PWR; BWR
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Source
Jan 1983; 59 p; Available from NTIS, PC A04/MF A01 as DE83901589
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Report
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Mager, T.R.; McLoughlin, V.J.
Westinghouse Electric Corp., Pittsburgh, Pa. (USA)1971
Westinghouse Electric Corp., Pittsburgh, Pa. (USA)1971
AbstractAbstract
No abstract available
Primary Subject
Source
Oct 1971; 47 p; HSSTP-TR--16; For Oak Ridge National Lab., Tenn.
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Report
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Mager, T.R.; Rishel, R.D.
Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Technology Div1982
Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Technology Div1982
AbstractAbstract
[en] The feasibility and methodology for thermal annealing an embrittled reactor vessel was investigated with the primary goal being to develop an in-situ thermal annealing procedure which would maximize fracture toughness recovery, minimize re-exposure sensitivity, and minimize reactor downtime. The feasibility was assessed by evaluating whether plant design and system limitations (reactor vessel design, metallurgical, reactor vessel insulation, primary shield concrete, RCS piping and associated equipment support, and reactor vessel internals and fuel storage) would be exceeded by an imposed localized 8500F heat source. In addition, health physics problems were also considered. Finally, a conceptual thermal annealing apparatus was developed and a general thermal annealing procedure was established. The results of this work indicate that an anneal temperature of 8500F for a holding period of 168 hours, obtained by using resistance-type heating elements attached to a support apparatus, could be used in an effort to restore the fracture toughness properties of ferritic materials of the reactor vessel as required by 10CFR Part 50 Appendix G
Original Title
PWR; BWR
Primary Subject
Secondary Subject
Source
Jul 1982; 56 p; Available from NTIS., PC A04/MF A01 as DE82905826
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Report
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