[Federal Register Volume 65, Number 162 (Monday, August 21, 2000)]
[Notices]
[Pages 50722-50723]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-21230]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-412]


Pennsylvania Power Company, Ohio Edison Company, The Cleveland 
Electric Illuminating Company, The Toledo Edison Company, Firstenergy 
Nuclear Operating Company; Beaver Valley Power Station, Unit No. 2; 
Environmental Assessment and Finding of No Significant Impact

    The Nuclear Regulatory Commission (NRC) is considering issuance of 
an exemption from the requirements of title 10 of the Code of Federal 
Regulations (10 CFR) Sec. 50.60(a), and 10 CFR part 50, Appendix G, for 
Facility Operating License No. NPF-73, issued to FirstEnergy Nuclear 
Operating Company (the licensee), for operation of the Beaver Valley 
Power Station, Unit No. 2 (BVPS-2), located in Beaver County, 
Pennsylvania.

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 submittal dated 
June 17, 1999, that the NRC staff exempt BVPS-2 from the requirements 
of 10 CFR part 50, Sec. 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.
    Code Case N-640 permits the use of an alternate reference fracture 
toughness (KIC fracture toughness curve instead of the 
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. 
Therefore, an exemption is required in order to apply the Code Case. It 
should be noted that, although Code Case N-640 was incorporated into 
the ASME Code recently, an exemption is still required because the 
proposed P/T limits (excluding Code Case N-640) are based on the 1989 
edition of the ASME Code.
    The proposed action is in accordance with the licensee's 
application for exemption dated June 17, 1999.

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 part 50, Sec. 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 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 15 
effective full power years based on ASME Code, Section XI, Appendix G, 
requirements. The results indicate that 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-2 
with P/T curves developed to satisfy ASME Code, Section XI, Appendix G,

[[Page 50723]]

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. Implementation of the proposed P/T curves, as 
allowed by ASME Code Case N-640, does not significantly reduce the 
margin of safety.
    In the associated exemption, the NRC 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 ASME 
Code Case N-640.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and 
concludes that the proposed action provides adequate margin of safety 
against brittle failure of the reactor coolant pressure boundary.
    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 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.

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-2.

Agencies and Persons Consulted

    In accordance with its stated policy, on July 10, 2000, 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 17, 1999, which is available for public 
inspection at the Commission's Public Document Room, The Gelman 
Building, 2120 L Street, NW., Washington, DC. 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).

    Dated at Rockville, Maryland, this 16th day of August, 2000.

    For the Nuclear Regulatory Commission.
Daniel S. Collins,
Project Manager, Section 1, Project Directorate I, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 00-21230 Filed 8-18-00; 8:45 am]
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