[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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