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Commonwealth Edison Company; Braidwood Station, Unit 2 Environmental Assessment and Finding Of No Significant Impact

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 [Federal Register: October 31, 1997 (Volume 62, Number 211)]
[Notices]
[Page 59008-59010]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr31oc97-116]

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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

 
 

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