[Federal Register Volume 61, Number 242 (Monday, December 16, 1996)]
[Notices]
[Pages 66062-66063]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-31813]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-266 and 50-301]
Wisconsin Electric Power Company; Point Beach Nuclear Plant,
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. DPR-24 and DPR-27,
issued to Wisconsin Electric Power Company, (the licensee), for
operation of Point Beach Nuclear Plant, Unit Nos. 1 and 2, located in
Manitowoc County, Wisconsin.
Environmental Assessment
Identification of the Proposed Action
The proposed action would allow the licensee to utilize the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code) Case N-514, ``Low Temperature Overpressure
Protection,'' to determine its low temperature overpressure protection
(LTOP) setpoints and is in accordance with the licensee's application
for exemption dated July 1, 1996, as supplemented November 18, 1996.
The proposed action requests an exemption from certain requirements of
10 CFR 50.60, ``Acceptance Criteria for Fracture Prevention Measures
for Lightwater Nuclear Power Reactors for Normal Operation,'' to allow
application of an alternate methodology to determine the LTOP setpoints
for Point Beach Nuclear Plant, Unit Nos. 1 and 2. The proposed
alternate methodology is consistent with guidelines developed by the
ASME Working Group 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 Code Case N-514 has been
incorporated into Appendix G of Section XI of the ASME Code and
published in the 1993 Addenda to Section XI. However, 10 CFR 50.55a,
``Codes and Standards,'' and Regulatory Guide 1.147, ``Inservice
Inspection Code Case Acceptability'' have not been updated to reflect
the acceptability of Code Case N-514.
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 Specifications 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.
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, which
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 10 CFR Part 50, Appendix G, may be used when an
exemption is granted by the Commission pursuant to 10 CFR 50.12.
To prevent transients that would produce excursions exceeding the
10 CFR Part 50, 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 below the
LTOP enabling temperature that would prevent the pressure in the
reactor vessel from exceeding the P/T limits of 10 CFR Part 50,
Appendix G. To prevent these valves from lifting as a result of normal
operating pressure surges (e.g., reactor coolant pump starting or
stopping) with the reactor coolant system in a water solid condition,
the operating pressure must be maintained below the PORV setpoint. The
licensee's current LTOP analysis indicates that using this 10 CFR Part
50, Appendix G, safety margin to determine the PORV setpoint requires
operation of the plant in a narrow range of pressure that could result
in the lifting of the PORVs during normal heatup and cooldown
operation. Using Code Case N-514 would allow the licensee to operate
without a restriction on the number of operating reactor coolant pumps
in the determination of the LTOP setpoint analysis. Therefore, the
licensee proposed that in determining the PORV setpoint for LTOP events
for Point Beach, the allowable pressure be determined using the safety
margins developed in an alternate methodology in lieu of the safety
margins required by 10 CFR Part 50, Appendix G. The alternate
methodology is consistent with ASME Code Case N-514. The content of
Code Case N-514 was incorporated into Appendix G of Section XI of the
ASME Code and published in the 1993 Addenda to Section XI.
An exemption from 10 CFR 50.60 is required to use the alternate
methodology for calculating the maximum allowable pressure for LTOP
considerations. By application dated July 1, 1996, as supplemented
November 18, 1996, the licensee requested an exemption from 10 CFR
50.60 to allow it to utilize the alternate methodology of Code Case N-
514 to compute its LTOP setpoints.
Environmental Impacts of the Proposed Action
Appendix G of the ASME Code requires that the P/T limits be
calculated (a) using a safety factor of 2 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 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 Point Beach reactor
vessel material.
In determining the PORV setpoint for LTOP events, the licensee
proposed the use of safety margins based on an alternate methodology
consistent with the proposed ASME Code Case N-514 which 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 of
the existing ASME Appendix G. This
[[Page 66063]]
results in a safety factor of 1.8 on pressure. All other factors,
including assumed flaw size and fracture toughness, remain the same.
Although this methodology would reduce the safety factor on pressure,
it was demonstrated in the Bases of ASME Code Case N-514 that due to
the isothermal nature of LTOP events, the margins with respect to
toughness for LTOP transients is within the range provided by ASME,
Section XI, Appendix G, for normal heatup and cooldown in the low
temperature range. Thus, applying Code Case N-514 will satisfy the
underlying purpose of 10 CFR 50.60 for fracture toughness requirements.
Further, by relieving the operational restrictions, the potential for
undesirable lifting of the PORV would be reduced, thereby improving
plant safety.
The change will not increase the probability or consequences of
accidents, no changes are being made in the types of any effluents that
may be released offsite, and there is no significant increase in the
allowable individual or cumulative occupational radiation exposure.
Accordingly, the Commission concludes that there are no significant
radiological environmental impacts associated with the proposed action.
With regard to potential nonradiological impacts, the proposed
action does involve features located entirely within the restricted
area as defined in 10 CFR Part 20. It does not affect nonradiological
plant effluents and has no other environmental impact. Accordingly, the
Commission concludes that there are no significant nonradiological
environmental impacts associated with the proposed 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 Point
Beach Nuclear Plant, Unit Nos. 1 and 2.
Agencies and Persons Consulted
In accordance with its stated policy, on November 29, 1996, the
staff consulted with the Wisconsin State official, Ms. Sarah Jenkins,
of the Public Service Commission of Wisconsin, 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 letters dated July 1 and November 18, 1996, 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 at the Joseph P. Mann Library,
1516 Sixteenth Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 10th day of December 1996.
For the Nuclear Regulatory Commission.
Linda L. Gundrum,
Project Manager, Project Directorate III-1, Division of Reactor
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 96-31813 Filed 12-13-96; 8:45 am]
BILLING CODE 7590-01-P
