40554 Federal Register / Vol. 62, No. 145 / Tuesday, July 29, 1997 / Notices

action and this alternative action are

similar.

Alternative Use of Resources

This exemption does not involve the

use of any resources not previously

considered in the Final Environmental

Statement related to the Catawba

Nuclear Station.

Agencies and Persons Contacted

In accordance with its stated policy,

on July 7, 1997, the staff consulted with

the South Carolina State official, Virgil

Autrey of the Bureau of Radiological

Health, South Carolina Department of

Health and Environmental Control,

regarding the environmental impact of

the proposed exemption. The State

official had no comments.

Finding of No Significant Impact

Based upon the foregoing

environmental assessment, the

Commission concludes that the

proposed exemption 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 the

proposed action, see the licensee’s

request for the exemption dated

February 4, 1997, as supplemented by

letter dated March 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 at the

York County Library, 138 East Black

Street, Rock Hill, South Carolina.

Dated at Rockville, Maryland, this 22nd

day of July 1997.

For the Nuclear Regulatory Commission.

Herbert N. Kerkow,

Director, Project Directorate II–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 97–19933 Filed 7–28–97; 8:45 am]

BILLING CODE 7590–01–P

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
Federal Register / Vol. 62, No. 145 / Tuesday, July 29, 1997 / Notices 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

NUCLEAR REGULATORY

COMMISSION

Sunshine Meeting

DATE: Weeks of July 28, August 4, 11,

and 18, 1997.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED:

Week of July 28

There are no meetings scheduled for

the week of July 28.

Week of August 4—Tentative

Monday, August 4

2:00 p.m. Briefing by International

Programs (Closed—Ex. 1)

3:00 p.m. Briefing on Investigative

Matters (Closed—Ex. 5 & 7)

Wednesday, August 6

9:30 a.m. Meeting with Northeast

Nuclear on Millstone (Public

Meeting) Contact: Bill Travers, 301–

415–1200)

2:00 p.m. Briefing on Shutdown Risk

Proposed Rule for Nuclear Power

Plants (Public Meeting) (Contact:

Tim Collins, 301–415–2897)

3:30 p.m. Affirmation Session (Public

Meeting) if needed)

Thursday, August 7

9:30 a.m. Meeting with NRC Executive

Council (Public Meeting) Contact:

James L. Blaha, 301–415–1703)

Week of August 11—Tentative

There are no meetings scheduled for

the week of August 11.

Week of August 18—Tentative

Friday, August 22

11:30 a.m. Affirmation Session (Public

Meeting) (if needed)

Note: The schedule for Commission

Meetings is subject to change on short notice.