[Federal Register Volume 63, Number 9 (Wednesday, January 14, 1998)]
[Notices]
[Pages 2268-2270]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-891]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457]


Commonwealth Edison Company; Byron Station, Units 1 and 2, and 
Braidwood Station, Units 1 and 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 to Facility Operating License Nos. NPF-37, NPF-66, NPF-72 
and NPF-77, issued to Commonwealth Edison Company (the licensee), for 
operation of Byron Station, Units 1 and 2, and Braidwood Station, Units 
1 and 2, located in Ogle County and Will County, Illinois, 
respectively.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would permit the licensee to use the 1996 
Addenda to the American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel Code (Code), Section XI, Appendix G, to determine 
the reactor vessel pressure-temperature (P-T) limits and the low-
temperature overpressure protection (LTOP) system setpoints. By 
application dated April 3, 1997, as supplemented by letter dated June 
19, 1997, the licensee requested an exemption from certain requirements 
of 10 CFR part 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 P-T limits and LTOP system setpoints for Byron, Units 1 
and 2, and Braidwood, Units 1 and 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 
1996 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 1996 Addenda to the ASME Code.

[[Page 2269]]

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. 10 CFR 
Part 50, Appendix G, 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, 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. 10 CFR 50.55a requires that any reference to 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. 10 CFR 50.60(b) specifies 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. The LTOP system includes pressure 
relieving devices called power-operated relief valves (PORVs) that are 
set to open at reduced pressure when reactor pressure and temperature 
are reduced. The PORVs prevent the pressure in the reactor vessel from 
exceeding the P-T limits. However, to prevent the PORVs 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, when instrument uncertainty is considered, the operating 
window between the PORV setpoint and the minimum pressure required for 
reactor coolant pump seals is small and presents difficulties for plant 
operation.
    To prevent pressure from exceeding the 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 1996 Addenda to the ASME Code, 
Section XI, Appendix G. These addenda to the Code would permit a 
slightly higher pressure inside the reactor and 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 flow 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 Byron/
Braidwood reactor vessel material.
    For determining the P-T limits, ComEd proposed to use the safety 
margins based on the 1996 Addenda to the ASME Code in lieu of the 1989 
Edition. When compared to the 1989 Edition of the ASME Code, the 1996 
Addenda permits the use of a lower stress intensity factor for 
determining the applied stress intensity due to pressure and thermal 
stresses. This results in a slight reduction in the applied stress 
intensity and a corresponding shift in the allowable pressure at a 
given temperature in the non-conservative direction; however, this 
difference is minor when compared to the explicit conservatism 
incorporated into the Code, and the changes in the stress intensity 
factor are supported by the work performed by J.A. Keeney and T.L. 
Dickson at Oak Ridge National Laboratory (ORNL) for the NRC, and 
others.
    1996 Addenda to the ASME Code require that the system pressure is 
maintained below the P-T limits during normal operation, but allows the 
pressure that may occur during LTOP events 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 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 the 1996 Addenda to the ASME 
Code, Section XI, Appendix G. 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 
that are developed using the 1996 Addenda to the ASME Code, Section XI, 
Appendix G, methodologies described above. 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 1996 Addenda to the ASME Code, Section XI, Appendix G, 
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 stress safety factor that may 
occur during LTOP events. Reduced operational challenges will reduce 
the potential for undesirable impacts to the environment.
    It should be noted that the provision to set the PORV setpoint such 
that it protects 110 percent of the P-T limits is already part of the 
Byron and Braidwood licensing basis. This provision was approved in the 
exemption to 10 CFR 50.60 granted to Byron on November 29, 1996, and to 
Braidwood on July 13, 1995, and December 12, 1997, for Units 1 and 2, 
respectively, to allow the use of ASME Code Case N-514. Therefore, 
while it represents a change from the 1989 Edition of the ASME Code, it 
is not a change to the licensing basis for these facilities.

Environmental Impacts of the Proposed Action

    The Commission has completed its review of the proposed action and 
concludes that the proposed action involves features located entirely 
within the protected areas 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.

[[Page 2270]]

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 
Byron Station or the Braidwood Station.

Agencies and Persons Consulted

    In accordance with its stated policy, on January 9, 1998, 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 April 3, 1997, as supplemented by letter dated 
June 19, 1997, which are available for public inspection at the 
Commission's Public Document Room, The Gelman Building, 2120 L Street, 
NW., Washington, DC, and at the Local Public Document Room located: For 
Byron, the Byron Public Library District, 109 N. Franklin, P.O. Box 
434, Byron, Illinois 61010; for Braidwood, the Wilmington Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

    Dated at Rockville, Maryland, this 9th day of January 1998.

    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. 98-891 Filed 1-13-98; 8:45 am]
BILLING CODE 7590-01-P