[Federal Register Volume 67, Number 33 (Tuesday, February 19, 2002)]
[Notices]
[Pages 7405-7406]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-3897]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-334]


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

    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

[[Page 7406]]

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

    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 
[email protected]. 

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