[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
