[Federal Register Volume 63, Number 82 (Wednesday, April 29, 1998)]
[Notices]
[Pages 23477-23478]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-11339]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-282 and 50-306]


Northern States Power Company (Prairie Island Nuclear Generating 
Plant, Units 1 and 2); Environmental Assessment and Finding of No 
Significant Impact

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an exemption from certain requirements of its 
regulations to Facility Operating License Nos. DPR-42 and DPR-60, 
issued to Northern States Power Company (NSP or the licensee), for 
operation of Prairie Island Nuclear Generating Plant, Units 1 and 2, 
located in Goodhue County, Minnesota.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would permit the licensee to use American 
Society of Mechanical Engineers (ASME) Code Case N-514 for setting the 
pressure setpoint of each unit's overpressure protection system (OPPS) 
so that the pressure-temperature (P-T) limits required by 10 CFR Part 
50, Appendix G, could be exceeded by 10 percent during a low 
temperature pressure transient. By application dated March 6, 1998, the 
licensee requested an exemption from certain requirements of 10 CFR 
50.60, ``Acceptance Criteria for Fracture Prevention Measures for 
Lightwater Nuclear Power Reactors for Normal Operation,'' and 10 CFR 
Part 50, Appendix G, ``Fracture Toughness Requirements.''

The Need for the Proposed Action

    Pursuant to 10 CFR 50.60, all lightwater nuclear power reactors 
must meet the fracture toughness requirements for the reactor coolant 
pressure boundary as set forth in 10 CFR Part 50, Appendix G. Appendix 
G of 10 CFR Part 50 defines P-T limits during any condition of normal 
operation, including anticipated operational occurrences and system 
hydrostatic tests to which the pressure boundary may be subjected over 
its service lifetime, and specifies that these P-T limits must be at 
least as conservative as the limits obtained by following the methods 
of analysis and the margins of safety of the ASME Code, Section XI, 
Appendix G.
    By letter dated March 6, 1998, NSP submitted an exemption request 
to enable the use of ASME Code Case N-514 as an alternative method for 
determining the OPPS setpoint. NSP

[[Page 23478]]

determined that the exemption request from the provisions of 10 CFR 
50.60 and Appendix G was necessary since these regulations require, as 
noted above, that the reactor vessel conditions not exceed the P-T 
limits established by Appendix G. In referring to 10 CFR 50.12 on 
specific exemptions, NSP cited special circumstances as stated in 10 
CFR 50.12(a)(2)(ii) on achieving the underlying purpose of the 
regulations as its basis for requesting this exemption.
    The underlying purpose of 10 CFR Part 50, Appendix G, is to 
establish fracture toughness requirements for the reactor coolant 
system (RCS) pressure boundary to provide adequate margins of safety 
during any condition of normal operation. The OPPS provides a physical 
means of protecting these limits. NSP proposed that establishing the 
OPPS pressure setpoint per the N-514 provisions such that the vessel 
pressure would not exceed 110 percent of the P-T limit allowables would 
still provide an acceptable level of safety and mitigate the potential 
for an inadvertent actuation of the OPPS.
    The plant operators must operate the plant in a pressure window 
that is between the minimum pressure required to preserve reactor 
coolant pump seals and at a maximum pressure that does not challenge 
the power-operated relief valve setpoint. Without the application of 
ASME Code Case N-514, Prairie Island would have an operating window 
that is too narrow to permit reasonable system makeup and pressure 
control. Further reduction of the OPPS setpoint below 500 psig would 
increase the probability that the reactor coolant pumps' no. 1 seal 
will fail as a result of OPPS operation, and that such a seal failure 
could produce a breach in the RCS boundary that could not be isolated. 
Therefore, inadvertent OPPS actuation could lead to a small break loss-
of-coolant accident and the unnecessary release of reactor coolant 
inside containment.

Environmental Impacts of the Proposed Action

    The Commission has completed its evaluation of the proposed action 
and concludes that the proposed action involves features located 
entirely within the protected areas as defined in 10 CFR Part 20.
    The proposed action will not 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 the allowable individual or cumulative occupational 
radiation exposure. Accordingly, the Commission concludes that there 
are no significant radiological environmental impacts associated with 
the proposed action.
    With regard to potential nonradiological impacts, the proposed 
action does involve features located entirely within the restricted 
area as defined in 10 CFR Part 20. It does not affect nonradiological 
plant effluents and has no other environmental impact. Accordingly, the 
Commission concludes that there are no significant nonradiological 
environmental impacts associated with the proposed action.

Alternative to the Proposed Action

    Since the Commission has concluded there is no measurable 
environmental impact associated with the proposed action, any 
alternatives with equal or greater environmental impact need not be 
evaluated. As an alternative to the proposed action, the staff 
considered denial of the proposed action. 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 the 
Prairie Island Nuclear Generating Plant.

Agencies ad Persons Consulted

    In accordance with its stated policy, on April 7, 1998, the staff 
consulted with the Minnesota State official, Mike McCarthy of the 
Department of Public Service, regarding the environmental impact of the 
proposed action. The state official had no comments.

Finding of No Significant Impact

    Based upon the environmental assessment, the Commission 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 March 6, 1998, which is available for public 
inspection at the Commission's Public Document Room, The Gelman 
Building, 2120 L Street, NW., Washington, DC, and at the Local Public 
Document Room located at the Minneapolis Public Library, Technology and 
Science Department, 300 Nicollet Mall, Minneapolis, Minnesota 55401.

    Dated at Rockville, Maryland, this 23rd day of April, 1998.

    For the Nuclear Regulatory Commission.
Cynthia A. Carpenter,
Director, Project Directorate III-1, Division of Reactor Projects--III/
IV, Office of Nuclear Reactor Regulation.
[FR Doc. 98-11339 Filed 4-28-98; 8:45 am]
BILLING CODE 7590-01-P