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ASTM Selected Technical Papers
Effects of Radiation on Materials: 22nd Symposium
By
TR Allen
TR Allen
1
Symposium Chair and Editor
?
University of Wisconsin
?
Madison, Wisconsin
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RG Lott
RG Lott
2
Symposium Co-Chair and Editor
?
Westinghouse Electric Company
?
Pittsburgh, Pennsylvania
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JT Busby
JT Busby
3
Symposium Co-Chair and Editor
?
Oak Ridge National Laboratory
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AS Kumar
AS Kumar
4
Symposium Co-Chair and Editor
?
University of Missouri-Rolla
?
Rolla, Missouri
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ISBN-10:
0-8031-3401-0
ISBN:
978-0-8031-3401-0
No. of Pages:
406
Publisher:
ASTM International
Publication date:
2006

The mechanistically-guided embrittlement correlation model adopted in ASTM E 900-02 was based on US, surveillance results current through calendar year 1998. There now exists an extensive amount of new surveillance data that includes a large amount of boiling water reactor (BWR) results from a supplemental surveillance program designed to augment the plant-specific BWR surveillance programs. These recent data allow a statistical check of the ASTM E 900-02 embrittlement correlation, as well as the NRC correlation model currently being used in the pressurized thermal shock (PTS) re-evaluation effort and the older Regulatory Guide 1.99, Revision 2 correlation. Even though the ASTM E 900-02 embrittlement correlation is a simplified version of the NRC model, a comparison of the two embrittlement correlation models utilizing the new database has proven to be revealing. Based on the new BWR data, both models are inadequate in their ability to predict BWR results. Other aspects of the two models are presented, as well as plans to develop a revised embrittlement correlation.

1.
Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Revision 2
,
Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission
,
Washington, DC
,
1988
.
2.
Eason
,
E. D.
,
Wright
,
J. E.
, and
Odette
,
G. R.
,
Improved Embrittlement Correlations for Reactor Pressure Vessel Steels
, NUREG/CR-6551,
U.S. Nuclear Regulatory Commission
, Washington, DC,
1998
.
3.
Kirk
,
M.
, “
Revision of AT30 Embrittlement Trend Curves
,” presented at the
EPRI MRP/NRC PTS Re-Evaluation
meeting in
Rockville, MD
,
2000
.
4.
Server
,
W.
,
English
,
C.
,
Naiman
,
D.
, and
Rosinski
,
S.
,
Charpy Embrittlement Correlations — Status of Combined Mechanistic and Statistical Bases for U.S. Pressure Vessel Steels (MRP-45)
, EPRI 1000705, Palo Alto, CA,
2001
.
5.
English
,
C.
,
Server
,
W.
, and
Lott
,
R.
,
Materials Reliability Program: Validation and Use of ASTMe 900-02 for Reactor Pressure Vessel Integrity (MRP-86)
, EPRI 1003537, Palo Alto, CA,
2003
.
6.
Carter
,
R. G.
,
Griesbach
,
T. J.
, and
Hardin
,
T. C.
, “
THE BWRVIP Integrated Surveillance Program
,”
2004 ASME Pressure Vessels and Piping Conference
,
ASME
,
New York
.
7.
Carter
,
R. G.
,
Soneda
,
N.
,
Dohi
,
K.
,
Hyde
,
J. M.
,
English
,
C. A.
, and
Server
,
W. L.
, “
Microstructural Characterization of Irradiation-Induced Cu-Enriched Clusters in Reactor Pressure Vessel Steels
,”
Journal of Nuclear Materials
 0022-3115, Vol.
298
,
2001
, p. 211.
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