10 C.F.R. § 50 app G to Part 50
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. [DELTA]RTNDT 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 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.
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 pressure1 | Requirements for pressure-temperature limits | Minimum temperature requirements |
| 1. Hydrostatic pressure and leak tests (core is not critical): | |||
| 1.a Fuel in the vessel | [LESS THAN EQUAL TO]20% | ASME Appendix G Limits | (2) |
| 1.b Fuel in the vessel | [GREATER THAN]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 | [LESS THAN EQUAL TO]20% | ASME Appendix G Limits | (2) |
| 2.b Core not critical | [GREATER THAN]20% | ASME Appendix G Limits | (2) + 120 °F (6) |
| 2.c Core critical | [LESS THAN EQUAL TO]20% | ASME Appendix G Limits + 40 °F | Larger of [(4)] or [(2) + 40 °F] |
| 2.d Core critical | [GREATER THAN]20% | ASME Appendix G Limits + 40 °F | Larger of [(4)] or [(2) + 160 °F] |
| 2.e Core critical for BWR (5) | [LESS THAN EQUAL TO]20% | ASME Appendix G Limits + 40 °F | (2) + 60 °F |
1 Percent of the preservice system hydrostatic test pressure.
2 The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload.
3 The highest reference temperature of the vessel.
4 The minimum permissible temperature for the inservice system hydrostatic pressure test.
5 For boiling water reactors (BWR) with water level within the normal range for power operation.
6 Lower temperatures are permissible if they can be justified by showing that the margins of safety of the controlling region are equivalent to those required for the beltline when it is controlling.
10 C.F.R. 50 app G to Part 50