[Federal Register Volume 63, Number 73 (Thursday, April 16, 1998)]
[Notices]
[Pages 18942-18943]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-10102]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-286]


In the Matter of Power Authority of the State of New York; 
(Indian Point Nuclear Generating Unit No. 3); Exemption

I

    The Power Authority of the State of New York (the licensee) is the 
holder of Facility Operating License No. DPR-64, which authorizes 
operation of the Indian Point Nuclear Generating Unit No. 3 (IP3). The 
license provides that the licensee is subject to all rules, 
regulations, and orders of the U.S. Nuclear Regulatory Commission (the 
Commission) now or hereafter in effect.
    The facility is a pressurized water reactor located in Westchester 
County, New York.

II

    The Code of Federal Regulations 10 CFR 50.60, states that the 
reactor coolant pressure boundaries for light water reactors must meet 
the fracture toughness and material surveillance program requirements 
set forth in Appendices G and H to 10 CFR Part 50.
    By letter dated January 28, 1998, the licensee requested an 
exemption from 10 CFR 50.60 to allow the use of an alternate 
methodology for the development of pressure-temperature (P-T) curves. 
As an alternative, the licensee proposed to use a methodology by ABB 
Combustion Engineering Nuclear Operations (the CE methodology).
    References in 10 CFR 50.60 and Appendix G require the use of a 
methodology at least as conservative as that found in Appendix G to the 
1989 Edition of Section XI of the ASME Code (the 1989 ASME Appendix G 
methodology or the 1989 methodology); therefore, the staff must review 
and approve the use of the CE methodology. The staff has reviewed the 
licensee's request and approves the use of the CE methodology in place 
of the 1989 methodology for the construction of reactor vessel 
pressure-temperature (P-T) limits as described in 10 CFR Part 50, 
Appendix G. The CE methodology was used in the licensee's P-T limit 
amendment submittal dated February 27, 1998.

III

    The NRC has established requirements in 10 CFR Part 50 to protect 
the integrity of the reactor coolant system pressure boundary. As a 
part of these, 10 CFR Part 50, Appendix

[[Page 18943]]

G requires that P-T limits be established for reactor pressure vessels 
(RPVs) during normal operation and vessel hydrostatic testing. In 
particular, 10 CFR Part 50, Appendix G, Section IV.2.b., requires that 
these 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.'' The Code of Federal Regulations at 
10 CFR 50.55(a) specifies that the applicable ASME Code is the 1989 
Edition. The Code of Federal Regulations at 10 CFR 50.60, which broadly 
addresses the establishment of criteria for fracture prevention, states 
that ``proposed alternatives to the described requirements in 
Appendices G and H of this part or portions thereof may be used when an 
exemption is granted by the Commission under Sec. 50.12.'' The licensee 
used the CE methodology for constructing its P-T limits in place of the 
1989 ASME Appendix G methodology approved by the staff in the 
regulations; therefore, the licensee applied for an exemption to use 
the CE methodology.

IV

    In the submittal, the exemption was requested under the special 
circumstances given in 10 CFR 50.12(a)(2)(ii). The provisions of this 
section state that special circumstances are present whenever 
``Application of the regulation in the particular circumstances would 
not serve the underlying purpose of the rule or is not necessary to 
achieve the underlying purpose of the rule.'' In the application, the 
licensee stated that ``The use of ABB-CE alternate methodology 
requested by this exemption provides greater operational flexibility 
while still maintaining reactor vessel integrity. In addition, the use 
of the ABB-CE methodology to generate pressure-temperature curves 
yields comparable results to the use of the ASME Appendix G 
methodology. Therefore, the reactor vessel is protected against 
nonductile failure and the underlying purpose of the rule is 
achieved.''
    The staff reviewed the licensee's application and the CE 
methodology and has concluded that this alternative method meets the 
underlying intent of the regulations. The thermal analysis method of 
the CE methodology consists of a plant-specific thermal analysis and a 
fracture mechanics analysis based on influence coefficients from finite 
element analyses under thermal loading. The staff review determined 
that this thermal analysis method is more rigorous than that of the 
1989 methodology and that the rest of the CE methodology is the same as 
the 1989 ASME Appendix G methodology. The staff concludes, therefore, 
that an exemption under the special circumstances of 10 CFR 
50.12(a)(2)(ii) is appropriate, and that the application of the CE 
methodology meets the underlying intent of the regulations.

V

    For the foregoing reasons, the NRC staff has concluded that the 
licensee's proposed use of the alternative methodology in determining 
the P-T limits will not present an undue risk to public health and 
safety and is consistent with the common defense and security. The NRC 
staff has determined that there are special circumstances present, as 
specified in 10 CFR 50.12(a)(2)(ii), in that application of 10 CFR 
50.60 is not necessary in order to achieve the underlying purpose of 
this regulation.
    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12, this exemption is authorized by law, will not present an undue 
risk to public health and safety, and is consistent with the common 
defense and security.
    Therefore, the Commission hereby grants the following exemption:
    The Power Authority of the State of New York is exempt from the 
requirements of 10 CFR 50.60 in that they are permitted to use the CE 
methodology detailed in their application for exemption dated January 
28, 1998, for developing P-T limits for the Indian Point Nuclear 
Generating Station Unit No. 3.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will have no significant impact on the 
quality of the human environment (63 FR 17902).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 10th day of April 1998.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 98-10102 Filed 4-15-98; 8:45 am]
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