The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an exemption from certain requirements of 10 CFR part 50, appendix G, for Facility Operating License No. DPR-49, issued to Nuclear Management Company, LLC (NMC, or the licensee) for operation of the Duane Arnold Energy Center (DAEC), located in Linn County, Iowa.
Identification of the Proposed Action
Title 10 of the Code of Federal Regulations (10 CFR part 50), appendix G, requires that pressure-temperature (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 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, Appendix G Limits.
To address provisions of amendments to the technical specifications (TS) P-T limits, the licensee requested in its submittal dated October 16, 2000, that the staff exempt NMC from application of specific requirements of 10 CFR part 50, appendix G, and substitute use of ASME Code Case N-640. The license amendment request is being addressed as a separate action. Code Case N-640 permits the use of an alternate reference fracture toughness (KIc fracture toughness curve instead of KIa fracture toughness curve) for reactor vessel materials in determining the P-T limits. Since the KIc fracture toughness curve shown in ASME 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 Section XI, Appendix G, Figure G-2210-1 (the KIa fracture toughness curve), using Code Case N-640 for establishing the P-T limits would be less conservative than the methodology currently endorsed by 10 CFR part 50, appendix G and, therefore, Start Printed Page 20693an exemption to apply the Code Case would be required by 10 CFR 50.60(b).
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 reactor coolant system (RCS) P-T limits, because continued use of the present curves unnecessarily restricts the P-T operating window. Since the RCS P-T operating window is defined by the P-T operating and test limit curves developed in accordance with the ASME Section XI, Appendix G procedure, continued operation of DAEC 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 degrees Fahrenheit in a limited operating window during the 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, does 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.
In the associated exemption, the staff has determined that, pursuant to 10 CFR 50.12(a)(2)(ii), the underlying purpose of the regulation will continue to be served by the implementation of this Code Case.
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 any effluents that may be released offsite, 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 environmental impacts, the proposed action does not involve any historic sites. It does not affect nonradiological plant effluents and has no other environmental impact. Therefore, there are no significant nonradiological impacts associated with the proposed action.
Accordingly, the NRC concludes that there are no significant environmental impacts associated with the proposed action.
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
This action does not involve the use of any resources not previously considered in the “Final Environmental Statement Relating to the Operation of the Duane Arnold Energy Center,” dated March 1973.
Agencies and Persons Consulted
In accordance with its stated policy, on March 26, 2001, the staff consulted with the Iowa State official, Mr. D. McGhee of the Department of Public Health, 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 letter dated October 16, 2000. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).Start Signature
Dated at Rockville, Maryland, this 17th day of April 2001.
For the Nuclear Regulatory Commission.
Carl F. Lyon,
Project Manager, Section 1, Project Directorate III, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-10095 Filed 4-23-01; 8:45 am]
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