[Federal Register Volume 62, Number 102 (Wednesday, May 28, 1997)]
[Notices]
[Pages 28907-28908]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-13867]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-302]


Florida Power Corporation; 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 No. DPR-72 issued to Florida 
Power Corporation, (the licensee), for operation of the Crystal River 
Unit 3 Nuclear Generating Plant (CR3) located in Citrus County, 
Florida.

Environmental Assessment

Identification of Proposed Action

    The proposed action is in accordance with the licensee's 
application dated April 7, 1997 for exemption from certain requirements 
of 10 CFR 50.60, ``Acceptance Criteria for Fracture Prevention Measures 
for Lightwater Nuclear Power Reactors for Normal Operation'' which 
would allow the licensee to utilize the American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-514, ``Low 
Temperature Overpressure Protection,'' to determine its low temperature 
overpressure protection (LTOP) setpoints. The licensee requests an 
exemption from certain requirements of 10 CFR 50.60, to allow 
application of an alternate methodology to determine the LTOP setpoints 
for CR3. The proposed alternate methodology is consistent with 
guidelines developed by the ASME Working Group to define pressure 
limits during LTOP events that avoid certain unnecessary operational 
restrictions, provide adequate margins against failure of the reactor 
pressure vessel, and reduce the potential for unnecessary activation of 
pressure-relieving devices used for LTOP. These guidelines have been 
incorporated into Code Case N-514, ``Low Temperature Overpressure 
Protection,'' which has been approved by the ASME Code Committee. The 
content of Code Case N-514 has been incorporated into Appendix G of 
Section XI of the ASME Code and published in the 1993 Addenda to 
Section XI. However, 10 CFR 50.55a, ``Codes and Standards,'' and 
Regulatory Guide 1.147, ``Inservice Inspection Code Case 
Acceptability'' have not been updated to reflect the acceptability of 
Code Case N-514.
    The philosophy used to develop Code Case N-514 guidelines is to 
ensure that the LTOP limits are still below the pressure/temperature 
(P/T) limits for normal operation but allow the pressure that may occur 
with activation of pressure-relieving devices to exceed the P/T limits, 
provided acceptable margins are maintained during these events. This 
philosophy protects the pressure vessel from LTOP events and still 
maintains the Technical Specifications P/T limits applicable for normal 
heatup and cooldown in accordance with 10 CFR part 50, Appendix G, and 
Sections III and XI of the ASME Code.

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, which 
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. 
It is specified in 10 CFR 50.60(b) that alternatives to the described 
requirements in 10 CFR part 50, Appendix G, may be used when an 
exemption is granted by the Commission pursuant to 10 CFR 50.12.
    To prevent transients that would produce excursions exceeding the 
10 CFR part 50, Appendix G, P/T limits while the reactor is operating 
at low temperatures, the licensee installed an LTOP system. The LTOP 
system includes a pressure-relieving device in the form of a power-
operated relief valve (PORV). The PORV is set at a pressure below the 
LTOP enabling temperature that would prevent the pressure in the 
reactor vessel from exceeding the P/T limits of 10 CFR part 50, 
Appendix G. To prevent the PORV from lifting as a result of normal 
operating pressure surges (e.g., reactor coolant pump starting or 
stopping) with the reactor coolant system in a water solid condition, 
the operating pressure must be maintained below the PORV setpoint. The 
licensee indicates that its LTOP PORV setpoint based on the 10 CFR part 
50, Appendix G, would restrict the P/T operating window and could 
potentially result in undesired actuation of the PORV during normal 
heatup and cooldown operation. The operating window is restricted by 
the difference between the P/T limit curves and the

[[Page 28908]]

reactor coolant pump net positive suction head curve. Therefore, the 
licensee proposed to use the safety margins developed in an alternate 
methodology in lieu of the safety margins required by 10 CFR part 50, 
Appendix G for determining the allowable pressure, and the PORV 
setpoint for LTOP events. The alternate methodology is consistent with 
ASME Code Case N-514. The content of Code Case N-514 was incorporated 
into Appendix G of Section XI of the ASME Code and published in the 
1993 Addenda to Section XI.
    An exemption from 10 CFR 50.60 is required to use the alternate 
methodology for calculating the maximum allowable pressure for LTOP 
considerations. By application dated April 7, 1997, the licensee 
requested an exemption from 10 CFR 50.60 to allow it to utilize the 
alternate methodology of Code Case N-514 for computing its LTOP 
setpoints.

Environmental Impacts of the Proposed Action

    Appendix G of the ASME Code requires that the P/T limits be 
calculated (a) Using a safety factor of 2 on the principal membrane 
(pressure) stresses, (b) assuming a flaw at the surface with a depth of 
one-quarter (\1/4\) of the vessel wall thickness and a length of 6 
times its depth, and (c) using a conservative fracture toughness curve 
that is based on the lower bound of static, dynamic, and crack arrest 
fracture toughness tests on material similar to the CR3 reactor vessel 
material.
    In determining the PORV setpoint for LTOP events, the licensee 
proposed the use of safety margins based on an alternate methodology 
consistent with the proposed ASME Code Case N-514, which allows 
determination of the setpoint for LTOP events such that the maximum 
pressure in the vessel will not exceed 110 percent of the P/T limits of 
the existing ASME Appendix G. All other factors, including assumed flaw 
size and fracture toughness, will be consistent with the 10 CFR 50.60, 
Appendix G. Although this methodology would reduce the safety factor on 
pressure, the margins with respect to toughness are acceptable for LTOP 
transients. Thus, applying Code Case N-514 will satisfy the underlying 
purpose of 10 CFR 50.60 for fracture toughness requirements. Further, 
by relieving the operational restrictions, the potential for 
undesirable lifting of the PORV would be reduced, thereby improving 
plant safety.
    The change 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

    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 did not involve the use of any resources not previously 
considered in the Final Environmental Statements related to operation 
of CR3, dated May 1973.

Agencies and Persons Consulted

    In accordance with its stated policy, on May 12, 1997 the staff 
consulted with the Florida State Official, Mr. Bill Passetti of the 
Florida Department of Health and Rehabilitative Services, regarding the 
environmental impact of the proposed action. The State official had no 
comments.

Finding of No Significant Impact

    The Commission has determined not to prepare an environmental 
impact statement for the proposed exemption. Based upon the foregoing 
environmental assessment, the Commission concludes that the proposed 
action will not have a significant effect on the quality of the human 
environment.
    For further details with respect to this action, see the request 
for exemption dated April 7, 1997 which is available for public 
inspection at the Commission's Public Document Room, 2120 L Street, 
NW., Washington, DC and at the local public document room located at 
Coastal Region Library, 8619 W. Crystal Street, Crystal River, Florida 
32629.

    Dated at Rockville, Maryland, this 21st day of May 1997.

    For the Nuclear Regulatory Commission.
Frederick J. Hebdon,
Director, Project Directorate II-3, Division of Reactor Projects--I/II, 
Office of Nuclear Reactor Regulation.
[FR Doc. 97-13867 Filed 5-27-97; 8:45 am]
BILLING CODE 7590-01-P