[Federal Register Volume 59, Number 197 (Thursday, October 13, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-25301]


[[Page Unknown]]

[Federal Register: October 13, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-369 and 50-370]

 

In the Matter of Duke Power Company (McGuire Nuclear Station 
Units 1 and 2); Exemption

I

    The Duke Power Company (the licensee) is the holder of Facility 
Operating License Nos. NPR-9 and NPF-17, which authorize operation of 
the McGuire Nuclear Station, Units 1 and 2, respectively. The licenses 
provide, among other things, that the licensee is subject to all rules, 
regulations, and orders of the Commission now or hereafter in effect.
    The facilities consist of two pressurized water reactors, McGuire 
Nuclear Station, Units 1 and 2, at the licensee's site located in 
Mecklenburg County, North Carolina.

II

    Title 10 CFR 50.60, ``Acceptance Criteria for Fracture Prevention 
Measures for Light-water Nuclear Power Reactors for Normal Operation,'' 
states that all light-water 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. Apendix 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. 10 CFR 50.60(b) specifies that alternatives to the described 
requirements in appendices G and H to 10 CFR part 50 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 appendix G P/T limits while 
the reactor is operating at low temperatures, the licensee installed a 
low temperature overpressure (LTOP) system. The system includes 
pressure-relieving devices called Power-Operated Relief Valves (PORVs). 
The PORVs are 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 appendix G P/T limits. To prevent the 
PORVs from lifting as a result of normal operating pressure surges 
(e.g., reactor coolant pump starting, and shifting operating charging 
pumps) with the reactor coolant system in a water solid condition, the 
operating pressure must be maintained below the PORV setpoint. In 
addition, in order to prevent cavitation of a reactor coolant pump, the 
operator must maintain a differential pressure across the reactor 
coolant pump seals. Hence, the licensee must operate the plant in a 
pressure window that is defined as the difference between the minimum 
required pressure to start a reactor coolant pump and the operating 
margin to prevent lifting of the PORVs due to normal operating pressure 
surges. The licensee LTOP analysis indicates that using the appendix G 
safety margins to determine the PORV setpoint would result in a 
pressure setpoint within its operating window, but there would be no 
margin for normal operating pressure surges. Therefore, operating with 
these limits could result in the lifting of the PORVs and cavitation of 
the reactor coolant pumps during normal operation.
    The licensee proposed that in determining the design setpoint for 
LTOP events for McGuire Units 1 and 2, the allowable pressure be 
determined using the safety margins developed in an alternate 
methodology in lieu of the safety margins currently required by 
appendix G, 10 CFR part 50. Designated Code Case N-514, the proposed 
alternate methodology is consistent with guidelines developed by the 
American Society of Mechanical Engineers (ASME) Working Group on 
Operating Plant Criteria 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. Code Case N-514, ``Low Temperature Overpressure 
Protection,'' 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. The 
MRC staff is revising 10 CFR 50.55a, which will endorse the 1993 
Addenda and appendix G of Section XI into the regulations.
    An exemption from 10 CFR 50.60 is required to use the alternate 
methodology for calculating the maximum allowable pressure for the LTOP 
setpoint. By application dated June 28, 1994, the licensee requested an 
exemption from 10 CFR 50.60 for this purpose.
    By letter dated August 18, 1994 (and further clarified by letter 
dated September 7, 1994), the licensee supplied additional information 
that described the use of a secondary side heat source to permit the 
heatup of the reactor coolant system, assuming that the exemption was 
not granted. Since the secondary side heat source could cause primary 
side transients, the staff considers the use of a secondary side heat 
source to be an undesirable method of operation.

III

    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 circumstance 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.
    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 six (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 McGuire 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 proposed 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 appendix G. This results in a safety factor 
of 1.8 on the principal membrane stresses. All other factors, including 
assumed flaw size and fracture toughness, remain the same. Although 
this methodology would reduce the safety factor on the principal 
membrane stresses, the proposed criteria will provide adequate margins 
of safety to the reactor vessel during LTOP transients and will satisfy 
the underlying purpose of 10 CFR 50.60 for fracture toughness 
requirements.
    Using the licensee's proposed safety factors instead of appendix G 
safety factors to calculate the LTOP setpoint will permit a higher LTOP 
setpoint than would otherwise be required and will provide added margin 
to prevent normal operating surges from lifting the PORVs or cavitation 
of the reactor coolant pumps.

IV

    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), such 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 the Duke Power 
Company 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 in order to be in 
compliance with these regulations. This exemption is applicable only to 
LTOP conditions during normal operation.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will not result in any significant adverse 
environmental impact. Publication of the Environmental Assessment and 
Finding of No Significant Impact in the Federal Register was delayed 
due to circumstances beyond the Commission's control. The Commission 
has determined that emergency circumstances exist and therefore is 
issuing this exemption pursuant to 10 CFR 51.13 and 51.35 without prior 
publication. Publication in the Federal Register will occur on October 
3, 1994.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 30th day of September, 1994.

    For the Nuclear Regulatory Commission.
John F. Stolz,
Acting Director, Division of Reactor Projects--I/II, Office of Nuclear 
Reactor Regulation.
[FR Doc. 94-25301 Filed 10-12-94 8:45 am]
BILLING CODE 7590-01-M