[Federal Register Volume 68, Number 136 (Wednesday, July 16, 2003)]
[Notices]
[Pages 42137-42139]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-17960]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-315]


Indiana Michigan Power Company Donald C. Cook Nuclear Plant, Unit 
1; 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-58, issued to Indiana Michigan Power Company (the licensee), 
for operation of the Donald C. Cook (D. C. Cook) Nuclear Plant, Unit 1, 
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 1, 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 
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 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

[[Page 42138]]

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, 
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 December 10, 2002.

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 based on the method specified by Appendix G to 10 CFR 
part 50, unnecessarily restrict 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 (OPPS) 
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 1, based on ASME Code, section XI, Appendix G 
requirements, would significantly restrict the ability to perform plant 
heatup and cooldown, create an unnecessary burden to plant operations, 
and challenge control of plant evolutions required with OPPS enabled. 
Continued operation of D. C. Cook Unit 1 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 protect system setpoints, as 
allowed by ASME Code Case N-641, will continue to provide significant 
safety margin for the RCPB.
    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. Nuclear Plant Units 1 and 2, dated August 1973.

Agencies and Persons Consulted

    On June 6, 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.

[[Page 42139]]

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 10, 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, 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 10th day of July 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. 03-17960 Filed 7-15-03; 8:45 am]
BILLING CODE 7590-01-P