[Federal Register Volume 62, Number 211 (Friday, October 31, 1997)]
[Notices]
[Pages 59008-59010]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-28881]


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NUCLEAR REGULATORY COMMISSION

[Docket No. STN 50-457]


Commonwealth Edison Company; Braidwood Station, Unit 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 for Facility Operating License No. NPF-77, issued to 
Commonwealth Edison Company, (ComEd, the licensee), for operation of 
the Braidwood Station, Unit 2, located in Will County, Illinois.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would permit the licensee to use the alternate 
methodology in American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel Code (Code) Case N-514, ``Low Temperature 
Overpressure Protection,'' to determine the low temperature 
overpressure protection (LTOP) system setpoints. By application dated 
November 30, 1994, as supplemented by letter dated May 11, 1995, 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.'' The exemption 
would allow application of an alternate methodology to determine the 
LTOP system setpoints for Braidwood, Unit 2. The proposed alternate 
methodology is consistent with guidelines developed by the 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. These guidelines have been incorporated into the 
1993 Addenda to the ASME Code, Section XI, Appendix G. However, 10 CFR 
50.55a, ``Codes and Standards,'' has not been updated to reflect the 
acceptability of the 1993 Addenda to 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

[[Page 59009]]

Part 50, Appendix G. Appendix G of 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, 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. It is required in 10 CFR 50.55a that any reference to the 
ASME Code, Section XI, in 10 CFR Part 50 refers to addenda through the 
1988 Addenda and editions through the 1989 Edition of the Code unless 
otherwise noted. 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 under 10 CFR 50.12.
    To prevent transients that would produce excursions exceeding the 
P-T limits while the reactor is operating at low temperatures, the 
licensee installed the LTOP system, which includes pressure relieving 
devices called power-operated relief valves (PORVs). The PORVs prevent 
the pressure in the reactor vessel from exceeding the P-T limits. 
However, to prevent the PORV from lifting as a result of normal 
operating pressure surges, some margin is needed between the normal 
operating pressure and the PORV setpoint. In addition, normal operating 
pressure must be high enough to prevent damage to reactor coolant pumps 
that may result from cavitation or inadequate differential pressure 
across the pump seals. Hence, the licensee must operate the plant in a 
pressure window that is defined as the difference between the minimum 
pressure required for reactor coolant pumps and the operating margin to 
prevent lifting of the PORVs. When instrument uncertainty is 
considered, the operating window is small and presents difficulties for 
plant operation.
    To meet the 10 CFR Part 50, Appendix G, P-T limits, the PORVs would 
be set to open at a pressure very close to the normal pressure inside 
the reactor. With the PORV setpoint close to the normal operating 
pressure, minor pressure perturbations that typically occur in the 
reactor could cause the PORVs to open. This is undesirable from the 
safety perspective because after every PORV opening there is some 
concern that the PORV may not reclose. A stuck open PORV would continue 
to discharge primary coolant and reduce reactor pressure until the 
discharge pathway was closed by operator action.
    The licensee requested use of the ASME Code Case N-514, ``Low 
Temperature Overpressure Protection,'' for the determination of the 
PORV setpoints. This code case would permit a slightly higher PORV 
setpoint during low-temperature shutdown conditions. This would reduce 
the likelihood for inadvertent opening of the PORVs.
    Appendix G of the ASME Code requires that the P-T limits be 
calculated: (a) using a safety factor of two 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 Braidwood 
reactor vessel material.
    ASME Code Case N-514 requires that the system pressure is 
maintained below the P-T limits during normal operation, but allows the 
pressure that may occur with the activation of pressure relieving 
devices (PORVs) to exceed the P-T limits, provided acceptable margins 
are maintained during these events. This approach protects the pressure 
vessel from LTOP events, and maintains the Technical Specification 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.
    In determining the PORV setpoint for LTOP events, the licensee 
proposed to use the safety margins of ASME Code Case N-514. This 
alternate methodology 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. This results in a safety factor of 1.8 on 
the principal membrane stresses. All other factors, including the 
assumed flaw size and fracture toughness, remain the same. Although 
this methodology would reduce the safety factor on the principal 
membrane stresses, use of the proposed criteria will provide adequate 
margins of safety for the reactor vessel during LTOP events.
    Use of the Code Case N-514 safety margins will reduce operational 
challenges during low temperature, low pressure operations. In terms of 
overall safety, the safety benefits derived from simplified operations 
and the reduced potential for undesirable opening of the PORVs will 
more than offset the reduction of the principal membrane safety factor. 
Reduced operational challenges will reduce the potential for 
undesirable impacts to the environment.

Environmental Impacts of the Proposed Action

    The proposed action involves features located entirely within the 
protected area as defined in 10 CFR Part 20.
    The proposed action will not result in an increase in the 
probability or consequences of accidents or result in a change in 
occupational or offsite dose. Therefore, there are no radiological 
impacts associated with the proposed action.
    The proposed action will not result in a change in nonradiological 
plant effluent and will have no other nonradiological environmental 
impact.
    Accordingly, the Commission concludes that there are no 
environmental impacts associated with this action.

Alternatives 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 
Braidwood Station.

Agencies and Persons Consulted

    In accordance with its stated policy, on October 22, 1997, the 
staff consulted with the Illinois State official, Frank Niziolek of the 
Illinois Department of Nuclear Safety, 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 November 30, 1994, as supplemented by letter 
dated May 11, 1995, which are available for public inspection at the 
Commission's Public Document Room, The Gelman Building, 2120 L Street, 
NW., Washington, DC,

[[Page 59010]]

and at the local public document room located at the Wilmington Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

    Dated at Rockville, Maryland, this 23rd day of October 1997.

    For the Nuclear Regulatory Commission.
George F. Dick, Jr.,
Senior Project Manager, Project Directorate III-2, Division of Reactor 
Projects III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 97-28881 Filed 10-30-97; 8:45 am]
BILLING CODE 7590-01-P