[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]
BILLING CODE 7590-01-P