[Federal Register Volume 68, Number 33 (Wednesday, February 19, 2003)]
[Notices]
[Pages 8052-8053]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-3936]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-263]


Nuclear Management Company, LLC; Monticello Nuclear Generating 
Plant; 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, ``Acceptance criteria for 
fracture prevention measures for light-water nuclear power reactors for 
normal operation,'' and 10 CFR Part 50, Appendix G, ``Fracture 
Toughness Requirements,'' for Facility Operating License No. DPR-22, 
issued to the Nuclear Management Company, LLC (the licensee), for 
operation of the Monticello Nuclear Generating Plant, located in Wright 
County, Minnesota. 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 exempt the licensee from the requirements 
of 10 CFR Part 50, Section 50.60(a) and Appendix G, which would allow 
the use of American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code (ASME Code) Code Case N-640 as the basis for revised 
reactor vessel pressure and temperature (P/T) limit curves in the 
Monticello Technical Specifications (TSs).
    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 
Appendices G and H to 10 CFR Part 50. Appendix G to 10 CFR Part 50 
requires that P/T limits be established for reactor pressure vessels 
(RPVs) during normal operating and hydrostatic or leak-rate testing 
conditions. Specifically, 10 CFR Part 50, Appendix G, states, ``The 
appropriate requirements on both the pressure-temperature limits and 
the minimum permissible temperature must be met for all conditions.'' 
Appendix G of 10 CFR Part 50 specifies that the requirements for these 
limits are the ASME Code, Section XI, Appendix G, 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 April 22, 2002, as supplemented by letter dated 
September 16, 2002.

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 curves 
unnecessarily restricts the P/T operating window. Since the P/T 
operating window is defined by the [chyph] P/T operating and test limit 
curves developed in accordance with the ASME Code, Section XI, Appendix 
G, procedure, continued operation of Monticello with these P/T curves 
without the relief provided by ASME Code Case N-640 would unnecessarily 
require the RPV to maintain a temperature exceeding 212 [deg]F in a 
limited operating window during the

[[Page 8053]]

pressure test. Consequently, steam vapor hazards would continue to be 
one of the safety concerns for personnel conducting inspections in 
primary containment. Implementation of the proposed P/T curves, as 
allowed by ASME Code Case N-640, would not significantly reduce the 
margin of safety and would eliminate steam vapor hazards by allowing 
inspections in primary containment to be conducted at a lower coolant 
temperature.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and 
concludes that there are no significant adverse environmental impacts 
associated with the proposed action.
    The proposed action will not significantly increase the probability 
or consequences of accidents, no changes are being made in the types of 
effluents that may be released off site, 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 NRC 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 
Monticello.

Agencies and Persons Consulted

    On February 11, 2003, the staff consulted with the Minnesota State 
official, Nancy Campbell of the Department of Commerce, regarding the 
environmental impact of the proposed action. The State official had no 
comments.

Finding of No Significant Impact

    On the basis of the 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 application dated April 22, 2002, as supplemented by letter 
dated September 16, 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 12th day of February 2003.

    For the Nuclear Regulatory Commission.
L. Raghavan,
Chief, Section 1, Project Directorate III, Division of Licensing 
Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 03-3936 Filed 2-18-03; 8:45 am]
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