[Federal Register Volume 60, Number 131 (Monday, July 10, 1995)]
[Notices]
[Pages 35570-35571]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-16809]



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NUCLEAR REGULATORY COMMISSION
[Docket No. STN 50-456]

Commonwealth Edison Company; Braidwood Station, Unit 1; 
Environmental Assessment and Finding of No Significant Impact
    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an exemption from Facility Operating License 
No. NPF-72, issued to the Commonwealth Edison Company (the licensee), 
for Braidwood Station, Unit 1, located in Will County, Illinois.

Environmental Assessment

Identification of Proposed Action

    The proposed action requests an exemption from certain requirements 
of 10 CFR 50.60, ``Acceptance Criteria for Fracture Prevention Measures 
for Light-Water Nuclear Power Reactors for Normal Operation,'' to allow 
application of an alternate methodology to determine the low 
temperature overpressure protection (LTOP) setpoint for Braidwood 
Station, Unit 1. The proposed alternate methodology is consistent with 
guidelines developed by the American Society of Mechanical Engineers 
(ASME) Working Group on Operating Plant Criteria (WGOPC) 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 Code Case N-514, ``Low 
Temperature Overpressure Protection,'' which 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 NRC staff is revising 10 CFR 50.55a, which will endorse 
the 1993 Addenda and Appendix G of Section XI into the regulations.
    The philosophy used to develop Code Case N-514 guidelines is to 
ensure that the LTOP limits are still below the pressure/temperature 
(P/T) limits for normal operation, but allow the pressure that may 
occur with activation of pressure-relieving devices to exceed the P/T 
limits, provided acceptable margins are maintained during these events. 
This philosophy protects the pressure vessel from LTOP events, and 
still maintains the Technical Specification P/T limits applicable for 
normal heatup and cooldown in accordance with Appendix G to 10 CFR Part 
50 and Sections III and XI of the ASME Code. The exemption was 
requested by the licensee by letter dated November 30, 1994, and 
supplemented by letter dated May 11, 1995.

The Need for the Proposed Action

    In 10 CFR 50.60 it 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. Appendix G to 10 CFR 50 
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. 
It is specified in 10 CFR 50.60(b) 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 transients that would produce pressure excursions 
exceeding the Appendix G P/T limits while the reactor is operating at 
low temperatures, the licensee installed an LTOP system. The LTOP 
system includes pressure relieving devices in the form of Power-
Operated Relief Valves (PORVs) that are set at a pressure low enough 
that if a transient occurred while the coolant temperature is below the 
LTOP enabling temperature, they would prevent the pressure in the 
reactor vessel from exceeding the Appendix G P/T limits. To prevent 
these valves 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's 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. Therefore, the licensee proposed that in determining the 
PORV setpoint for LTOP events for Braidwood, the allowable pressure be 
determined using the safety margins developed in an alternate 
methodology in lieu of the safety margins required by Appendix G to 10 
CFR Part 50. The alternate methodology is consistent with ASME Code 
Case N-514.
    An exemption from 10 CFR 50.60 is required to use the alternate 
methodology for calculating the maximum allowable pressure for LTOP 
considerations.

Environmemntal Impacts of the Proposed Action

    The Commission has completed its evaluation of the licensee's 
application.
    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.

[[Page 35571]]

    In determining the PORV setpoint for LTOP events, the licensee 
proposed to use safety margins based on an alternate methodology 
consistent with the proposed ASME Code 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, use of the proposed criteria will provide 
adequate margins of safety to the reactor vessel during LTOP 
transients.
    Accordingly, the Commission concludes that this proposed action 
would result in no significant radiological environmental impact.
    With regard to potential non-radiological impacts, the proposed 
change involves use of more realistic safety margins for determining 
the PORV setpoint during LTOP events. It does not affect non-
radiological plant effluents and has no other environmental impact. 
Therefore, the Commission concludes that there are no significant non-
radiological environmental impacts associated with the proposed 
exemption.

Alternative to the Proposed Action

    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 did not involve the use of any resources not previously 
considered in the Final Environmental Statements related to operation 
of Braidwood Station.

Agencies and Persons Consulted

    In accordance with its stated policy, on June 15, 1995, the staff 
consulted with the Illinois State Official, Mr. Frank Niziolek; Head, 
Reactor Safety Section; Division of Engineering; 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 foregoing 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 exemption.
    For further details with respect to this action, see the request 
for exemption dated November 30, 1994, as supplemented May 11, 1995, 
which is available for public inspection at the Commission's Public 
Document Room, 2120 L Street, NW., Washington, DC 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 3rd day of July 1995.

    For the Nuclear Regulatory Commission.
Ramin R. Assa,
Project Director, Project Directorate III-2, Division of Reactor 
Projects-III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 95-16809 Filed 7-6-95; 8:45 am]
BILLING CODE 7590-01-M