-
1
-
-
34648826603
-
-
ASME Code Case N-629, Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials Section XI Division 1, 1999.
-
-
-
-
2
-
-
34648821954
-
-
ASME, 2001 section III, NB-2331, Test requirements and acceptance standards, Materials for Vessels, 2001.
-
-
-
-
3
-
-
34648832861
-
-
ASTM, 2002a. E900-02, Standard guide for predicting radiation induced transition temperature shift in reactor vessel materials, E706 (IIF).
-
-
-
-
4
-
-
34648818148
-
-
ASTM, 2002b. E1921-02, Standard test method for determination of reference temperature, T0, for ferritic steels in the transition range.
-
-
-
-
5
-
-
85168671194
-
A comparison between French surveillance program results and predictions of irradiation embrittlement, Influence of radiation on material properties
-
Brillaud C., Hedin F., and Houssin B. A comparison between French surveillance program results and predictions of irradiation embrittlement, Influence of radiation on material properties. 13th International Symposium ASTM STP 956 (1987) 420-447
-
(1987)
13th International Symposium ASTM STP 956
, pp. 420-447
-
-
Brillaud, C.1
Hedin, F.2
Houssin, B.3
-
6
-
-
34648818882
-
-
Eason, E.D., Wright, J.E., Odette, G.R., 1998. Improved Embrittlement Correlations for Reactor Pressure Vessel Steels, NUREG/CR-6551.
-
-
-
-
7
-
-
32144434521
-
Radiation embrittlement of reactor pressure vessel steels
-
Elsevier Science (chapter 6.08)
-
English C.A., and Hyde J.M. Radiation embrittlement of reactor pressure vessel steels. Comprehensive Structural Integrity (2003), Elsevier Science (chapter 6.08)
-
(2003)
Comprehensive Structural Integrity
-
-
English, C.A.1
Hyde, J.M.2
-
8
-
-
34648821952
-
-
Gerard, R., 1995. Survey of National Regulatory Requirements, AMES Report No. 4, EUR 16305 EN.
-
-
-
-
9
-
-
34648832860
-
-
IAEA-TECDOC-1120, 1999. Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels.
-
-
-
-
10
-
-
34648818146
-
-
KTA 3201.2, 1996. Components of the Reactor Coolant Pressure Boundary on Light Water Reactors, Part 2: Design and Analysis.
-
-
-
-
11
-
-
34648832859
-
-
KTA 3203, 2001. Surveillance of the Irradiation Behaviour of Reactor Pressure Vessel Materials of LWR Facilities.
-
-
-
-
12
-
-
34648812659
-
-
Oosterkamp, W.J., Dufour, L.B., 1983. Neutron embrittlement of the reactor vessel in Borssele as determined from Charpy specimens, Kema scientific & technical reports, vol. 1, no. 4, pp. 45-54.
-
-
-
-
13
-
-
34648826601
-
-
Petrequin A., 1996. A review of formulas for predicting irradiation embrittlement of reactor vessel materials, AMES Report no. 6, EUR 16455 EN.
-
-
-
-
14
-
-
34648824360
-
-
Proposed Rule Making, 1996. 10 CFR Part 50. Section 50.60, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, NRC.
-
-
-
-
15
-
-
34648826600
-
-
Proposed Rule Making, 1984. 10 CFR Part 50. Section 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock, NRC.
-
-
-
-
16
-
-
34648818147
-
-
Proposed Rule Making, 2003. 10 CFR Part 50. Section 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock, NRC.
-
-
-
-
17
-
-
34648832858
-
-
Proposed Rule Making, 2003b. 10 CFR Part 50. Appendix G. Fracture toughness requirements, NRC.
-
-
-
-
18
-
-
34648821951
-
-
Proposed Rule Making 10 CFR part 50, 2003c. Appendix H, Reactor vessel material surveillance program requirements, NRC.
-
-
-
-
19
-
-
34648824361
-
-
RCC-M section III article MC1240, 1993a. Determination of reference nil ductility transition temperature.
-
-
-
-
20
-
-
34648814961
-
-
RCC-M Subsection Z, Appendix ZG 3430, 1993b. Irradiation effects.
-
-
-
-
21
-
-
34648818881
-
-
Reg. Guide 1.99, 1997. Radiation embrittlement of reactor vessel materials, rev. 1.
-
-
-
-
22
-
-
34648829789
-
-
Reg. Guide 1.99, 1988. Radiation embrittlement of reactor vessel materials, rev. 2.
-
-
-
-
23
-
-
34648812657
-
-
Regulatory Guide 1.154, 1987. Format and content of plant-specific pressurized thermal shock safety analysis reports for pressurized water reactors.
-
-
-
-
24
-
-
34648818143
-
-
RSEM, 1990. In service inspection rules for mechanical equipment of PWR nuclear islands (in French), article B-7212.
-
-
-
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