The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an exemption from certain requirements of 10 CFR 50.60(a) for Facility Operating Licenses Nos. NPF-11 and NPF-18, issued to Commonwealth Edison Company (ComEd, or the licensee) for operation of LaSalle County Station, Units 1 and 2, located in LaSalle County, Illinois.
Identification of the Proposed Action
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 February 29, 2000, that the staff exempt ComEd from application of specific requirements of 10 CFR Part 50, Section 50.60(a) and Appendix G, and substitute use of ASME Code Cases N-588 and N-640. Code Case N-588 permits the postulation of a circumferentially-oriented flaw (in lieu of an axially-oriented flaw) for the evaluation of the circumferential welds in RPV P-T limit curves. 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 pressure stresses on a circumferentially-oriented flaw are lower than the pressure stresses on an axially-oriented flaw by a factor of two, using Code Case N-588 for establishing the P-T limits would be less conservative than the methodology currently endorsed by 10 CFR Part 50, Appendix G and, therefore, an exemption to apply the Code Case would be required by 10 CFR 50.60(b). Likewise, 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, an exemption to apply the Code Case would also be required by 10 CFR 50.60(b).
The proposed action is in accordance with the licensee's application for exemption dated February 29, 2000.
The Need for the Proposed Action
The proposed exemption is needed to allow the licensee to implement ASME Code Case N-588 and 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 LaSalle 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 Cases N-588 and 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 Start Printed Page 60987served by the implementation of these Code Cases.
Environmental Impacts of the Proposed Action
The Commission 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 Commission 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 for LaSalle County Station, Units 1 and 2.
Agencies and Persons Consulted
In accordance with its stated policy, on July 19, 2000, the staff consulted with the Illinois State official, Frank Niziolek of the Illinois Department of Nuclear Safety, 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 Commission 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 February 29, 2000, which is available for public inspection 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 5th day of October 2000.
For the Nuclear Regulatory Commission.
Anthony J. Mendiola,
Chief, Section 2, Project Directorate III, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 00-26339 Filed 10-12-00; 8:45 am]
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