[Federal Register Volume 62, Number 133 (Friday, July 11, 1997)]
[Notices]
[Pages 37316-37317]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-18211]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-302]


Florida Power Corporation, Crystal River Nuclear Generating Plant 
Unit 3; Exemption

I

    Florida Power Corporation (the licensee) is the holder of Facility 
Operating License No. DPR-72, which authorizes operation of the Crystal 
River Nuclear Generating Plant Unit 3 (CR3). The license provides, 
among other things, that the licensee is subject to all rules, 
regulations, and orders of the Commission now or hereafter in effect.
    The facility is of a pressurized water reactor type and is located 
in Citrus County, Florida.

II

    In its letter dated April 7, 1997, the licensee requested an 
exemption from the Commission's regulations. Title 10 of the Code of 
Federal Regulations, Part 50, Section 60 (10 CFR 50.60), ``Acceptance 
Criteria for Fracture Prevention Measures for Lightwater Nuclear Power 
Reactors for Normal Operation,'' states that all lightwater nuclear 
power reactors must meet the fracture toughness and material 
surveillance program requirements for the reactor coolant pressure 
boundary as set forth in Appendices G and H to 10 CFR Part 50. Appendix 
G to 10 CFR Part 50 defines pressure/temperature (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. Pursuant to 10 CFR 
50.60(b), alternatives to the Appendices G and H to 10 CFR Part 50 
requirements may be used when an exemption is granted by the Commission 
under 10 CFR 50.12.
    To prevent low-temperature overpressure transients that would 
produce pressure excursions exceeding the P/T limits of Appendix G to 
10 CFR Part 50 while the reactor is operating at low temperatures, the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code) Section XI requires that a low-temperature 
overpressure protection (LTOP) system shall be effective at coolant 
temperatures less than 200 deg.F or at coolant temperatures 
corresponding to a reactor vessel metal temperature less than reference 
temperature nil-ductility (RTNDT)+50 deg.F, whichever is 
greater.
    At CR3, the LTOP system includes a pressure-relieving device; 
power-operated relief valve (PORV). The PORV is to be set at a pressure 
low enough so that if an LTOP transient occurred, the mitigation system 
would prevent the pressure in the reactor vessel from exceeding the P/T 
limits of Appendix G to 10 CFR Part 50. To prevent the PORVs from 
lifting as a result of normal operating pressure surges (e.g., reactor 
coolant pumps starting or stopping) with the reactor coolant system 
(RCS) 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 reactor coolant pump net positive suction 
head curve.
    The licensee indicates that plant operation with this restriction 
places an unnecessary burden on plant operators to ensure safety limits 
are maintained, and could potentially result in an undesired actuation 
of the PORV during normal heatup and cooldown operation. Therefore, the 
licensee proposed that the PORV setpoint for LTOP events be determined 
using the safety margins developed in an alternate methodology in lieu 
of the safety margins required by 10 CFR Part 50, Appendix G. The 
alternate methodology would be consistent with ASME Code Case N-514, 
``Low Temperature Overpressure Protection,'' which allows exceeding the 
pressure of the P/T limits of 10 CFR Part 50, Appendix G, by 10 
percent. ASME Code Case N-514 is consistent with guidelines developed 
by the ASME Working Group on Operating Plant Criteria to define 
pressure limits during LTOP events. The code case methodology is 
intended to 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. ASME Code Case N-514 has been approved by the 
ASME Code Committee. The content of this code case has been 
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.

III

    Presently, CR3 Technical Specifications (TS) do not include LTOP 
features. By letter dated June 7, 1997, the licensee confirmed lowering 
the PORV setpoint to 454 psig. These values are based on the approved 
15 effective full power years (EFPY) P/T curves for normal cooldown and 
heatup, using the methodology described in ASME Code, Appendix G, with 
no reactor coolant pumps running. The licensee also confirmed that it 
currently controls LTOP features administratively using operating 
procedures (OPs). These OPs:
    (1) Limit the Pressurizer level to less than 220 inches to 
accommodate a water level surge, and RCS pressure to 100 psig,
    (2) Require both trains of High Pressure Injection (HPI) valves to 
be closed and breakers secured to prevent

[[Page 37317]]

inadvertent HPI into the RCS during LTOP conditions, and
    (3) Require the Core Flood Tank (CFT) pressure to be maintained 
within maximum allowable RCS P/T when CFT isolation valves are open, or 
these valves are closed to prevent inadvertent CFT injection into the 
RCS.
    The licensee stated that these administrative controls will remain 
in effect until the TS are revised to include LTOP features addressing 
the full range of RCS pressures.

IV

    Pursuant to 10 CFR 50.12, the Commission may, upon application by 
any interested person or upon its own initiative, grant exemptions from 
the requirements of 10 CFR Part 50 when (1) The exemptions are 
authorized by law, will not present an undue risk to public health or 
safety, and are consistent with the common defense and security and (2) 
when special circumstances are present. Special circumstances are 
present whenever, according to 10 CFR 50.12(a)(2)(ii), ``Application of 
the regulation in the particular circumstances would not serve the 
underlying purpose of the rule or is not necessary to achieve the 
underlying purpose of the rule.''
    The underlying purpose of 10 CFR 50.60, Appendix G, is to establish 
fracture toughness requirements for ferritic materials of pressure-
retaining components of the reactor coolant pressure boundary to 
provide adequate margins of safety during any condition of normal 
operation, including anticipated operational occurrences, to which the 
pressure boundary may be subjected over its service lifetime. Section 
IV.A.2 of this appendix requires that the reactor vessel be operated 
with P/T limits at least as conservative as those obtained by following 
the methods of analysis and the required margins of safety of Appendix 
G of the ASME Code, Section XI.
    Appendix G of Section XI 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 Point Beach 
reactor vessel material.
    In determining the setpoint for LTOP events, the licensee proposed 
to use safety margins based on an alternate methodology consistent with 
the ASME Code Case N-514 guidelines. The ASME Code Case N-514 allows 
determination of the setpoint for LTOP events such that the maximum 
pressure in the vessel would not exceed 110 percent of the P/T limits 
of the existing ASME Code, Section XI, Appendix G. This approach 
results in a safety factor of 1.8 on pressure. All other factors, 
including assumed flaw size and fracture toughness, remain the same. 
Although this methodology would reduce the safety factor on pressure, 
the margin with respect to toughness for LTOP transients are 
acceptable. 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.

V

    For the foregoing reasons, the NRC staff has concluded that the 
licensee's proposed use of the alternate methodology in determining the 
acceptable setpoint for LTOP events will not present an undue risk to 
public health and safety and is consistent with the common defense and 
security. The NRC staff has determined that there are special 
circumstances present, as specified in 10 CFR 50.12(a)(2)(ii), in that 
application of 10 CFR 50.60 is not necessary in order to achieve the 
underlying purpose of this regulation.
    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), an exemption is authorized by law, will not endanger life or 
property or common defense and security, and is otherwise in the public 
interest. Therefore, the Commission hereby grants an exemption from the 
requirements of 10 CFR 50.60 such that in determining the setpoint for 
LTOP events, the Appendix G curves for P/T limits are not exceeded by 
more than 10 percent.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will not have a significant effect on the 
quality of the human environment (62 FR 28907).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 3rd day of July 1997.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-18211 Filed 7-10-97; 8:45 am]
BILLING CODE 7590-01-P