Skip to Content

Notice

Pennsylvania Power Company, Ohio Edison Company, FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Unit No. 1; Environmental Assessment and Finding of No Significant Impact

Document Details

Information about this document as published in the Federal Register.

Published Document

This document has been published in the Federal Register. Use the PDF linked in the document sidebar for the official electronic format.

Start Preamble

The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an exemption from the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.60(a), and 10 CFR part 50, Appendix G, for Facility Operating License No. DPR-66, issued to FirstEnergy Nuclear Operating Company (the licensee), for operation of the Beaver Valley Power Station, Unit No. 1 (BVPS-1), located in Beaver County, Pennsylvania. 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

Appendix G to 10 CFR part 50 requires that pressure/temperature (P/T) limits be established for reactor pressure vessels during normal operating and hydrostatic or leak rate testing conditions. Specifically, this regulation states, “The appropriate requirements on both the pressure-temperature limits and the minimum permissible temperature must be met for all conditions.” Additionally, it specifies that the requirements for these limits are contained in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, Appendix G.

To address provisions of an amendment to the Technical Specification P/T limits, the licensee requested in its application dated June 29, 2001, as supplemented by letters of October 4 and December 1, 2001, that the NRC staff exempt BVPS-1 from the requirements of 10 CFR, Section 50.60(a), and 10 CFR Part 50, Appendix G, to allow application of ASME Code Case N-640 in establishing the reactor vessel pressure limits at low temperatures.

ASME Code Case N-640 permits the use of an alternate reference fracture toughness (Kc fracture toughness curve instead of the Ka fracture toughness curve) for reactor vessel materials in determining the P/T limits. Since the Kc fracture toughness curve shown in ASME Code, Section XI, Appendix A, Figure A-2200-1 (the Kc fracture toughness curve), provides greater allowable fracture toughness than the corresponding Ka fracture toughness curve of ASME Code, Section XI, Appendix G, Figure G-2210-1 (the Ka 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. Therefore, an exemption is required in order to apply ASME Code Case N-640.

The proposed action is in accordance with the licensee's application for exemption dated June 29, 2001, and supplements dated October 4 and December 1, 2001.

The Need for the Proposed Action

ASME Code Case N-640 is needed to revise the method used to determine the reactor coolant system (RCS) P/T limits.

The purpose of 10 CFR 50.60(a), and 10 CFR part 50, Appendix G, is to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. This protection is accomplished through these regulations that, in part, specify fracture toughness requirements for ferritic materials of the reactor coolant pressure boundary. Pursuant to 10 CFR part 50, Appendix G, it is required that P/T limits for the RCS be at least as conservative as those obtained by applying the methodology of the ASME Code, Section XI, Appendix G.

Current overpressure protection system (OPPS) setpoints 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 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. The impact on the P/T limits and OPPS setpoints has been evaluated for an increased service period to 22 effective full power years based on ASME Code, Section XI, Appendix G, requirements. The results indicate that the OPPS 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 BVPS-1 with P/T curves developed to satisfy ASME Code, Section XI, Appendix G, requirements without the relief provided by ASME Code Case N-640 would unnecessarily restrict the P/T operating window, especially at low-temperature conditions.

Application of ASME Code Case N-640 will provide results which are sufficiently conservative to ensure the integrity of the reactor coolant pressure boundary while providing P/T curves which are not overly restrictive.

In the associated exemption, the NRC staff would determine 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 ASME Code Case N-640.

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 ASME Code Case N-640 to develop the new P-T limits and OPPS setpoints.

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 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 involve any historic Start Printed Page 7406sites. 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

This action does not involve the use of any resources not previously considered in the Final Environmental Statement for BVPS-1 dated July 1973.

Agencies and Persons Consulted

On January 24, 2002, the staff consulted with the Pennsylvania State official, Mr. L. Ryan, of the Pennsylvania Department of Environmental Protection Bureau, Division 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 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 June 29, 2001, as supplemented by letters dated October 4 and December 1, 2001. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, 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 pdr@nrc.gov.

Start Signature

Dated at Rockville, Maryland this 11th day of February 2002.

For the Nuclear Regulatory Commission.

Daniel Collins,

Project Manager, Section 1, Project Directorate I, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

End Signature End Preamble

[FR Doc. 02-3897 Filed 2-15-02; 8:45 am]

BILLING CODE 7590-01-P