[Federal Register Volume 62, Number 145 (Tuesday, July 29, 1997)]
[Notices]
[Pages 40554-40555]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-19931]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-244]
Rochester Gas and Electric Corporation; R. E. Ginna Nuclear Power
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 for Facility Operating License No. DRP-18 issued to
Rochester Gas and Electric Corporation (the licensee), for operation of
the R. E. Ginna Nuclear Power Plant located in Wayne County, New York.
Environmental Assessment
Identification of 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 June 12, 1997. 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 the R. E. Ginna Nuclear
Power Plant. The proposed alternate methodology is consistent with
guidelines developed by the 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
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. Appendix
G of 10 CFR part 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 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
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 (RCP) starting
and shifting operating charging pumps) with the reactor coolant system
in a solid water condition, the operating pressure must be maintained
below the PORV setpoint.
In addition, to prevent damage to RCP seals, the operator must
maintain a minimum differential pressure across the RCP 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 RCP and the operating margin to prevent lifting of the PORVs
due to normal operating pressure surges. 10 CFR part 50, Appendix G,
safety margin adds instrument uncertainty in the LTOP setpoint. The
licensee's current LTOP analysis indicates that using this 10 CFR part
50, Appendix G, safety margin to determine the PORV setpoint would
result in an operating window between the LTOP setpoint and the minimum
pressure required for RCP seals which is significantly restricted when
physical conditions such as PORV overshoot, RCP Ps, and static
head corrections are taken into account in setpoint determination.
Operating with these limits could result in the lifting of the PORVs or
damage to the RCP seals during normal operation. Using Code Case N-514
would allow the licensee to
[[Page 40555]]
recapture most of the operating margin that is lost by factoring in the
instrument uncertainties in the determination of the LTOP setpoint. The
net effect of using Code Case N-514 is that the setpoint will not
change significantly with the next setpoint analysis. Therefore, the
licensee proposed that in determining the setpoint for LTOP events for
Ginna, 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 the ASME Code Case N-514. The content of this Code Case
had been 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 June 12, 1997, 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 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 Ginna
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 guidelines. ASME Code
Case N-514 allows determination of the setpoint for LTOP events such
that the maximum pressure in the vessel will not exceed 110% 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.
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 the R.
E. Ginna Nuclear Power Plant'' dated December 1975.
Agencies and Persons Consulted
In accordance with its stated policy, on July 2, 1997, the staff
consulted with the Mr. Jack Spath of the New York State Energy Research
and Development Authority, 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 June 12, 1997, which is available for public
inspection at the Commission's Public Document Room, which is located
at The Gelman Building, 2120 L Street, NW., Washington, DC, and at the
local public document room located at the Rochester Public Library, 115
South Avenue, Rochester, New York.
Dated at Rockville, Maryland, this 23rd day of July 1997.
For the Nuclear Regulatory Commission.
Guy S. Vissing,
Senior Project Manager, Project Directorate I-1, Division of Reactor
Projects I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 97-19931 Filed 7-28-97; 8:45 am]
BILLING CODE 7590-01-P