[Federal Register Volume 68, Number 53 (Wednesday, March 19, 2003)]
[Notices]
[Pages 13336-13338]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-6544]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-316]
Indiana Michigan Power Company, Donald C. Cook Nuclear Plant,
Unit 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
No. DPR-74, issued to Indiana Michigan Power Company (the licensee),
for operation of the Donald C. Cook (D. C. Cook) Nuclear Plant, Unit 2,
located in Berrien County, Michigan. 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-641 as the basis for revised
reactor vessel pressure and temperature (P-T) curves, and low
temperature overpressure protection system setpoints in the D. C. Cook
Unit 2 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
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 P-T 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 (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,
[[Page 13337]]
respectively) for reactor vessel materials in determining the P-T
curves and low temperature overpressure protection system setpoints for
effective temperature and allowable pressure. Since the KIC
fracture toughness curve shown in ASME Code, section XI, Appendix A,
Figure A-2200-1 (the KIC fracture toughness curve), provides
greater allowable fracture toughness 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. Therefore, an exemption to apply ASME Code Case N-641 is
required.
The proposed action is in accordance with the licensee's
application dated July 23, 2002, as supplemented by letters dated
November 15, 2002, and January 24, 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 low temperature overpressure protection
system setpoints continued use of the method specified by Appendix G to
10 CFR part 50, 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 32 effective full-power years for D.C. Cook Unit
2, based on ASME Code, section XI, Appendix G requirements, would
significantly restrict the ability to perform plant heatup and
cooldown, and create an unnecessary burden to plant operations, and
challenge control of plant evolutions required with the Over Pressure
Protection Section enabled. Continued operation of D.C. Cook Unit 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 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 different resource than
those previously considered in the Final Environmental Statement for
the Donald C. Cook Nuclear Plant Units 1 and 2, dated August 1973.
Agencies and Persons Consulted
On February 10, 2003, the staff consulted with the Michigan State
official, Ms. Sara De Cair of the Department of Environmental Quality,
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
[[Page 13338]]
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 July 23, 2002, as supplemented by letters dated
November 15, 2002, and January 24, 2003. Documents may be examined,
and/or copied for a fee, at 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 12th day of March 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. 03-6544 Filed 3-18-03; 8:45 am]
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