Appendix G to Part 50 - Fracture Toughness Requirements
10:1.0.1.1.30.0.117.88.29 : Appendix G
Appendix G to Part 50 - Fracture Toughness Requirements I.
Introduction and scope. II. Definitions. III. Fracture toughness
tests. IV. Fracture toughness requirements. I. Introduction and
Scope
This appendix specifies fracture toughness requirements for
ferritic materials of pressure-retaining components of the reactor
coolant pressure boundary of light water nuclear power reactors to
provide adequate margins of safety during any condition of normal
operation, including anticipated operational occurrences and system
hydrostatic tests, to which the pressure boundary may be subjected
over its service lifetime.
The ASME Code forms the basis for the requirements of this
appendix. “ASME Code” means the American Society of Mechanical
Engineers Boiler and Pressure Vessel Code. If no section is
specified, the reference is to Section III, Division 1, “Rules for
Construction of Nuclear Power Plant Components.” “Section XI” means
Section XI, Division 1, “Rules for Inservice Inspection of Nuclear
Power Plant Components.” If no edition or addenda are specified,
the ASME Code edition and addenda and any limitations and
modifications thereof, which are specified in § 50.55a, are
applicable.
The sections, editions and addenda of the ASME Boiler and
Pressure Vessel Code specified in § 50.55a have been approved for
incorporation by reference by the Director of the Federal Register.
A notice of any changes made to the material incorporated by
reference will be published in the Federal Register. Copies of the
ASME Boiler and Pressure Vessel Code may be purchased from the
American Society of Mechanical Engineers, United Engineering
Center, 345 East 47th Street, New York, NY 10017, and are available
for inspection at the NRC Library, 11545 Rockville Pike, Two White
Flint North, Rockville, MD 20852-2738.
The requirements of this appendix apply to the following
materials:
A. Carbon and low-alloy ferritic steel plate, forgings,
castings, and pipe with specified minimum yield strengths not over
50,000 psi (345 MPa), and to those with specified minimum yield
strengths greater than 50,000 psi (345 MPa) but not over 90,000 psi
(621 MPa) if qualified by using methods equivalent to those
described in paragraph G-2110 of appendix G of section XI of the
latest edition and addenda of the ASME Code incorporated by
reference into § 50.55a(b)(2).
B. Welds and weld heat-affected zones in the materials specified
in paragraph I.A. of this appendix.
C. Materials for bolting and other types of fasteners with
specified minimum yield strengths not over 130,000 psi (896
MPa).
Note:
The adequacy of the fracture toughness of other ferritic
materials not covered in this section must be demonstrated to the
Director, Office of Nuclear Reactor Regulation, on an individual
case basis.
II. Definitions
A. Ferritic material means carbon and low-alloy steels,
higher alloy steels including all stainless alloys of the 4xx
series, and maraging and precipitation hardening steels with a
predominantly body-centered cubic crystal structure.
B. System hydrostatic tests means all preoperational
system leakage and hydrostatic pressure tests and all system
leakage and hydrostatic pressure tests performed during the service
life of the pressure boundary in compliance with the ASME Code,
Section XI.
C. Specified minimum yield strength means the minimum
yield strength (in the unirradiated condition) of a material
specified in the construction code under which the component is
built under § 50.55a.
D. RTNDT means the reference temperature of the material,
for all conditions.
(i) For the pre-service or unirradiated condition, RTNDT is
evaluated according to the procedures in the ASME Code, Paragraph
NB-2331.
(ii) For the reactor vessel beltline materials, RTNDT must
account for the effects of neutron radiation.
E. ΔRTNDT means the transition temperature shift, or
change in RTNDT, due to neutron radiation effects, which is
evaluated as the difference in the 30 ft-lb (41 J) index
temperatures from the average Charpy curves measured before and
after irradiation.
F. Beltline or Beltline region of reactor vessel
means the region of the reactor vessel (shell material including
welds, heat affected zones, and plates or forgings) that directly
surrounds the effective height of the active core and adjacent
regions of the reactor vessel that are predicted to experience
sufficient neutron radiation damage to be considered in the
selection of the most limiting material with regard to radiation
damage.
III. Fracture Toughness Tests
A. To demonstrate compliance with the fracture toughness
requirements of section IV of this appendix, ferritic materials
must be tested in accordance with the ASME Code and, for the
beltline materials, the test requirements of appendix H of this
part. For a reactor vessel that was constructed to an ASME code
earlier than the Summer 1972 Addenda of the 1971 Edition (under §
50.55a), the fracture toughness data and data analysis must be
supplemented in a manner approved by the Director, Office of
Nuclear Reactor Regulation, to demonstrate equivalence with the
fracture toughness requirements of this appendix.
B. Test methods for supplemental fracture toughness tests
described in paragraph IV.A.1.b of this appendix must be submitted
to and approved by the Director, Office of Nuclear Reactor
Regulation, prior to testing.
C. All fracture toughness test programs conducted in accordance
with paragraphs III.A and III.B must comply with ASME Code
requirements for calibration of test equipment, qualification of
test personnel, and retention of records of these functions and of
the test data.
IV. Fracture Toughness Requirements
A. The pressure-retaining components of the reactor coolant
pressure boundary that are made of ferritic materials must meet the
requirements of the ASME Code, supplemented by the additional
requirements set forth below, for fracture toughness during system
hydrostatic tests and any condition of normal operation, including
anticipated operational occurrences. Reactor vessels may continue
to be operated only for that service period within which the
requirements of this section are satisfied. For the reactor vessel
beltline materials, including welds, plates and forgings, the
values of RTNDT and Charpy upper-shelf energy must account for the
effects of neutron radiation, including the results of the
surveillance program of appendix H of this part. The effects of
neutron radiation must consider the radiation conditions
(i.e., the fluence) at the deepest point on the crack front
of the flaw assumed in the analysis.
1. Reactor Vessel Charpy Upper-Shelf Energy Requirements
a. Reactor vessel beltline materials must have Charpy
upper-shelf energy 1 in the transverse direction for base material
and along the weld for weld material according to the ASME Code, of
no less than 75 ft-lb (102 J) initially and must maintain Charpy
upper-shelf energy throughout the life of the vessel of no less
than 50 ft-lb (68 J), unless it is demonstrated in a manner
approved by the Director, Office of Nuclear Reactor Regulation,
that lower values of Charpy upper-shelf energy will provide margins
of safety against fracture equivalent to those required by Appendix
G of Section XI of the ASME Code. This analysis must use the latest
edition and addenda of the ASME Code incorporated by reference into
§ 50.55a(b)(2) at the time the analysis is submitted.
1 Defined in ASTME 185-79 and -82 which are incorporated by
reference in appendix H to part 50.
b. Additional evidence of the fracture toughness of the beltline
materials after exposure to neutron irradiation may be obtained
from results of supplemental fracture toughness tests for use in
the analysis specified in section IV.A.1.a.
c. The analysis for satisfying the requirements of section
IV.A.1 of this appendix must be submitted, as specified in § 50.4,
for review and approval on an individual case basis at least three
years prior to the date when the predicted Charpy upper-shelf
energy will no longer satisfy the requirements of section IV.A.1 of
this appendix, or on a schedule approved by the Director, Office of
Nuclear Reactor Regulation.
2. Pressure-Temperature Limits and Minimum Temperature Requirements
a. Pressure-temperature limits and minimum temperature
requirements for the reactor vessel are given in table 1, and are
defined by the operating condition (i.e., hydrostatic pressure and
leak tests, or normal operation including anticipated operational
occurrences), the vessel pressure, whether or not fuel is in the
vessel, and whether the core is critical. In table 1, the vessel
pressure is defined as a percentage of the preservice system
hydrostatic test pressure. The appropriate requirements on both the
pressure-temperature limits and the minimum permissible temperature
must be met for all conditions.
b. The pressure-temperature limits identified as “ASME Appendix
G limits” in table 1 require that the limits must be at least as
conservative as limits obtained by following the methods of
analysis and the margins of safety of Appendix G of Section XI of
the ASME Code.
c. The minimum temperature requirements given in table 1 pertain
to the controlling material, which is either the material in the
closure flange or the material in the beltline region with the
highest reference temperature. As specified in table 1, the minimum
temperature requirements and the controlling material depend on the
operating condition (i.e., hydrostatic pressure and leak tests, or
normal operation including anticipated operational occurrences),
the vessel pressure, whether fuel is in the vessel, and whether the
core is critical. The metal temperature of the controlling
material, in the region of the controlling material which has the
least favorable combination of stress and temperature must exceed
the appropriate minimum temperature requirement for the condition
and pressure of the vessel specified in table 1.
d. Pressure tests and leak tests of the reactor vessel that are
required by Section XI of the ASME Code must be completed before
the core is critical.
B. If the procedures of section IV.A. of this appendix do not
indicate the existence of an equivalent safety margin, the reactor
vessel beltline may be given a thermal annealing treatment to
recover the fracture toughness of the material, subject to the
requirements of § 50.66. The reactor vessel may continue to be
operated only for that service period within which the predicted
fracture toughness of the beltline region materials satisfies the
requirements of section IV.A. of this appendix using the values of
RTNDT and Charpy upper-shelf energy that include the effects of
annealing and subsequent irradiation.
Table 1 - Pressure and Temperature
Requirements for the Reactor Pressure Vessel
Operating condition |
Vessel pressure
1 |
Requirements for
pressure-temperature limits |
Minimum temperature
requirements |
1. Hydrostatic
pressure and leak tests (core is not critical): |
|
|
|
1.a Fuel in the
vessel |
≤20% |
ASME Appendix G Limits |
( 2) |
1.b Fuel in the
vessel |
>20% |
ASME Appendix G Limits |
( 2) + 90 °F (
6) |
1.c No fuel in
the vessel (Preservice Hydrotest Only) |
ALL |
(Not Applicable) |
( 3) + 60 °F |
2. Normal
operation (incl. heat-up and cool-down), including anticipated
operational occurrences: |
|
|
|
2.a Core not
critical |
≤20% |
ASME Appendix G Limits |
( 2) |
2.b Core not
critical |
>20% |
ASME Appendix G Limits |
( 2) + 120 °F (
6) |
2.c Core
critical |
≤20% |
ASME Appendix G Limits + 40
°F |
Larger of [( 4)] or
[( 2) + 40 °F] |
2.d Core
critical |
>20% |
ASME Appendix G Limits + 40
°F |
Larger of [( 4)] or
[( 2) + 160 °F] |
2.e Core
critical for BWR ( 5) |
≤20% |
ASME Appendix G Limits + 40
°F |
( 2) + 60 °F |
[60 FR 65474, Dec. 19, 1995, as amended at 73 FR 5723, Jan. 31,
2008; 78 FR 34248, June 7, 2013; 78 FR 75450, Dec. 12, 2013; 84 FR
65644, Nov. 29, 2019]