[Federal Register Volume 68, Number 102 (Wednesday, May 28, 2003)]
[Notices]
[Pages 31735-31736]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-13218]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-369 and 50-370]
Duke Power Company; McGuire Nuclear Station, Unit Nos. 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, Appendix G, for Facility Operating
License Nos. NPF-9 and NPF-17, issued to Duke Power Company (the
licensee), for operation of the McGuire Nuclear Station, Unit Nos. 1
and 2 (McGuire), nuclear power plant, located in Mecklenburg County,
North Carolina. 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, appendix G, which would allow the use of American
Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME
Code) Code Case N-641 as the basis for revised reactor vessel pressure
(RVP) and temperature (P/T) curves, and low temperature overpressure
protection system setpoints in the McGuire, Unit Nos. 1 and 2 Technical
Specifications.
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 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-641 permits the use of alternate reference
fracture toughness for reactor vessel materials in determining the P/T
curves and low temperature overpressure protection system setpoints for
effective temperature and allowable pressure. The alternate reference
fracture toughness involves the use of the ``KIC fracture
toughness curve'' instead of the ``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. Since the KIC fracture toughness curve
shown in ASME Code, Section XI, Appendix A, Figure A-2200-1 (the
KIC fracture toughness curve), provides a higher fracture
toughness value than the corresponding KIA fracture
toughness curve of ASME Code, Section XI, Appendix G, Figure G-2210-1
(the KIA fracture toughness curve), using ASME Code Case N-
641 to establish the P/T curves and low temperature overpressure
protection system setpoints would be less conservative than the
methodology currently endorsed by 10 CFR part 50, Appendix G. The
provisions of ASME Code Case N-641 were incorporated into the Appendix
G to Section XI of the ASME Code in the 1998 Edition through 2000
Addenda which is the Edition and Addenda of record in the 2003 edition
of 10 CFR part 50. However, in this case, the McGuire licensing basis
has only been updated to include the 1995 Edition through 1996 Addenda
of the ASME Code. tHerefore, an exemption to apply ASME Code Case N-641
is required.
The poposed action is in accordance with the licensee's application
dated December 12, 2002, as supplemented by letters dated March 27 and
April 23, 2003.
The Need for the Proposed Action
The proposed exemption is needed to allow the licensee to implement
ASME Code Case N-641 in order to revise the method used to determine
the P/T curves and because the continued use of the method specified by
Appendix G to 10 CFR part 50, to develop low temperature overpressure
protection system setpoints unnecessarily restricts the P/T operating
window.
The underlying purpose of Appendix G, is to protect the integrity
of the reactor coolant pressure boundary (RCPB) in nuclear power
plants. This is accomplished through regulations that, in part, specify
fracture toughness requirements for ferritic materials of the RCPB.
Pursuant to 10 CFR part 50, appendix G, it is required that P/T limits
for the reactor coolant system (RCS) be at least as conservative as
those obtained by applying the methodology of the ASME Code, Section
XI, Appendix G. Current P/T limits produce operational constraints by
limiting the P/T range available to the operator to heat up or cool
down the plant. The operating window through which the operator heats
up and cools down the RCS becomes more restrictive with continued
reactor vessel service. Reducing this operating window could
potentially have an adverse safety impact by increasing the possibility
of inadvertent low temperature overpressure protection system actuation
due to pressure surges associated with normal plant evolutions, such as
reactor coolant pump start and swapping operating charging pumps with
the RCS in a water-solid condition. P/T limits for an increased service
period of operation of 34 effective full-power years for McGuire, Unit
Nos. 1 and 2, based on ASME Code, Section XI, Appendix G requirements,
would significantly restrict the ability to perform plant heatup and
cooldown, and would create an unnecessary burden to plant operations,
and challenge control of plant evolutions required with the Over
Pressure Protection feature enabled. Continued operation of McGuire,
Unit Nos. 1 and 2, with P/T curves developed to satisfy ASME Code,
Section XI, Appendix G, requirements without the relief provided by
ASME Code Case N-641 would unnecessarily restrict the P/T operating
window, especially at low temperature conditions. Use of the
KIC curve in determining the lower bound fracture toughness
of RPV steels is more technically correct than use of the
KIA curve, since the rate of loading during a heatup or
cooldown is slow and is more representative of a static condition than
a dynamic condition. The KIC curve appropriately implements
the use of static initiation fracture toughness behavior to evaluate
the controlled heatup and cooldown process of a reactor vessel. The
staff has required use of the conservatism of the KIA curve
since 1974, when the curve was adopted by the ASME Code. This
conservatism was initially necessary due to the limited knowledge of
the fracture
[[Page 31736]]
toughness of RPV materials at that time. Since 1974, additional
knowledge has been gained about RPV materials, which demonstrates that
the lower bound on fracture toughness provided by the KIA
curve greatly exceeds the margin of safety required, and that the
KIC curve is sufficiently conservative to protect the public
health and safety from potential RPV failure. Application of ASME Code
Case N-641 will provide results that are sufficiently conservative to
ensure the integrity of the RCPB, while providing P/T curves and low
temperature overpressure protection system setpoints that are not
overly restrictive. Implementation of the proposed P/T curves and low
temperature overpressure protection system setpoints, as allowed by
ASME Code Case N-641, does not significantly reduce the margin of
safety.
In the associated exemption, the NRC staff has determined that,
pursuant to 10 CFR part 50, Section 50.12(a)(2)(ii), the underlying
purpose of the regulation will continue to be served by the
implementation of ASME Code Case N-641.
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 RCS P/T limits.
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 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 resources not previously
considered in NUREG-0063, ``Final Environmental Statement Related to
the Operation of William B. McGuire Nuclear Station, Units 1 and 2,''
April 1976 and the Addendum to NUREG-0063 issued in January 1981.
Agencies and Persons Consulted
In accordance with its stated policy, on May 19, 2003, the staff
consulted with the North Carolina State official, Mr. Johnny James of
the Division of Environmental Health, Radiation Protection Section,
North Carolina Department of Environment and Natural Resources,
regarding the environmental impact of the proposed amendments. 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 letter dated December 12, 2002, as supplemented by letters
dated March 27 and April 23, 2003. Documents may be examined, and/or
copied for a fee, a the NRC's Public Document Room (PDR), located at
One White Flint North, Public File Area O1 F21, 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 21st day of May 2003.
For the Nuclear Regulatory Commission.
John A. Nakoski,
Chief, Section 1, Project Directorate II, Division of Licensing Project
Management, Office of Nuclear Reactor Regulation.
[FR Doc. 03-13218 Filed 5-27-03; 8:45 am]
BILLING CODE 7590-01-P