[Federal Register Volume 68, Number 145 (Tuesday, July 29, 2003)]
[Notices]
[Pages 44550-44551]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-19213]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-327 and 50-328]


Tennessee Valley Authority; Sequoyah Nuclear Plant, Units 1 and 
2, Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering 
issuance of an exemption from title 10 of the Code of Federal 
Regulations (10 CFR) part 50, section 50.60 for Facility Operating 
License Nos. DPR-77 and DPR-79, issued to the Tennessee Valley 
Authority (TVA, the licensee), for operation of the Sequoyah Nuclear 
Plant (SQN), Units 1 and 2, located in Hamilton County, Tennessee. 
Therefore, as required by 10 CFR 51.21, the NRC is issuing this 
environmental assessment and finding of no significant impact.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would permit the use of the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 
Section XI Code Case N-640, ``Alternative Requirement Fracture 
Toughness for Development of P-T Limit Curves for ASME B&PV Code, 
Section XI, Code Case N-640,'' in lieu of 10 CFR 50, Appendix G, 
paragraph IV.A.2.b.
    The regulation at 10 CFR part 50, section 50.60(a), requires, in 
part, that except where an exemption is granted by the Commission, all 
light-water nuclear power reactors must meet the fracture toughness 
requirements for the reactor coolant pressure boundary set forth in 
Appendix G to 10 CFR part 50. Appendix G of 10 CFR part 50 requires the 
establishment of pressure-temperature (P-T) limits for specific 
material fracture toughness requirements of the reactor coolant 
pressure boundary materials and mandates the use of the ASME B&PV Code, 
Section XI, Appendix G. The requirements in 10 CFR 50, Appendix

[[Page 44551]]

G, establish an adequate margin to brittle failure during normal 
operation, anticipated operational occurrences, and system hydrostatic 
tests.
    ASME B&PV Code, Section XI, Code Case N-640 permits the use of an 
alternate reference fracture toughness curve for reactor pressure 
vessel materials for use in determining the P-T limits. ASME Code Case 
N-640 permits the use of alternate reference fracture toughness (i.e., 
use of ``KIC fracture toughness curve'' instead of 
``KIA fracture toughness curve,'' where KIC and 
KIA are ``Reference Stress Intensity Factors,'' as defined 
in ASME Code, Section XI, Appendices A and G, respectively) for reactor 
vessel materials in determining the P-T limits. Since the 
KIC fracture toughness curve shown in ASME Code, Section XI, 
Appendix A, Figure A-2200-1, provides greater allowable fracture 
toughness than the corresponding KIA fracture toughness 
curve of ASME Code, Section XI, Appendix G, Figure G-2210-1, using ASME 
Code Case N-640 to establish the P-T limits would be less conservative 
than the methodology currently endorsed by 10 CFR part 50, Appendix G. 
Therefore, an exemption to apply ASME Code Case N-640 is required.
    The proposed action is in accordance with the licensee's 
application dated September 6, 2002, as supplemented by letter dated 
December 19, 2002 and June 24, 2003.

The Need for the Proposed Action

    The proposed exemption is needed to allow the licensee to implement 
ASME Code Case N-640 in order to revise the method used to determine 
the P-T limits because continued use of the present method for 
determining P-T limits unnecessarily restricts the P-T operating 
window. The two primary benefits to the licensee from the use of Code 
Case N-640 are:
    [sbull] Challenges to the operators would be reduced since the 
requirements for maintaining high-vessel temperature during pressure 
testing would be lessened.
    [sbull] Enhanced personnel safety would result because of the lower 
temperatures which would exist during the conduct of inspections in 
primary containment.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and 
concludes that there are no significant environmental impacts 
associated with the use of the alternative analysis method to support 
the revision of the reactor coolant system P-T limits.
    The proposed action will not significantly increase the probability 
or consequences of accidents, no changes are being made in the types or 
significant increase in the amounts of effluents that may be released 
offsite, and there is no significant increase in occupational or public 
radiation exposure. Therefore, there are no significant radiological 
environmental impacts associated with the proposed action.
    With regard to potential nonradiological impacts, the proposed 
action does not have a potential to affect any historic sites. It does 
not affect nonradiological plant effluents and has no other 
environmental impact. Therefore, there are no significant 
nonradiological environmental impacts associated with the proposed 
action.
    Accordingly, the NRC concludes that there are no significant 
environmental impacts associated with the proposed action.

Environmental Impacts of the Alternatives to the Proposed Action

    As an alternative to the proposed action, the staff considered 
denial of the proposed action (i.e., the ``no-action'' alternative). 
Denial of the application would result in no change in current 
environmental impacts. The environmental impacts of the proposed action 
and the alternative action are similar.

Alternative Use of Resources

    The action does not involve the use of any different resource than 
those previously considered in the Final Environmental Statement for 
SQN, dated February 13, 1974.

Agencies and Persons Consulted

    On July 15, 2003, the staff consulted with the Tennessee State 
official, Ms. Elizabeth Flannagan, regarding the environmental impact 
of the proposed action. The State official had no comments.

Finding of No Significant Impact

    On the basis of this environmental assessment, the NRC concludes 
that the proposed action will not have a significant effect on the 
quality of the human environment. Accordingly, the NRC has determined 
not to prepare an environmental impact statement for the proposed 
action.
    For further details with respect to the proposed action, see the 
licensee's letter dated September 6, 2002, as supplemented by letter 
dated December 19, 2002. Documents may be examined, and/or copied for a 
fee, at the NRC's Public Document Room (PDR), located at One White 
Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible electronically from the 
Agencywide Documents Access and Management System (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC PDR Reference staff by telephone at 1-
800-397-4209 or 301-415-4737, or by e-mail to [email protected].

    Dated at Rockville, Maryland, this 23rd day of July 2003.

    For The Nuclear Regulatory Commission.
Allen G. Howe,
Chief, Section 2, Project Directorate 2, Division of Licensing Project 
Management, Office of Nuclear Reactor Regulation.
[FR Doc. 03-19213 Filed 7-28-03; 8:45 am]
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