Federal Register / Vol. 62, No. 145 / Tuesday, July 29, 1997 / Notices 40553

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 January 17, 1997, which is

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 rooms located at the

Salem Free Public Library, 112 West

Broadway, Salem, New Jersey, for Salem

and at the Pennsville Public Library,

190 S. Broadway, Pennsville, New

Jersey, for Hope Creek.

Dated at Rockville, Maryland, this 23rd day

of July 1997.

For the Nuclear Regulatory Commission.

John F. Stolz,

Director Project Directorate, I–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 97–19932 Filed 7–28–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–413 and 50–414]

Duke Power Company, et al.; Catawba

Nuclear 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–35 and NPF–52, issued to

Duke Power Company, et al. (the

licensee), for operation of the Catawba

Nuclear Station, Units 1 and 2, located

in York County, South Carolina.

Environmental Assessment

Identification of Proposed Action

The proposed action would exempt

the licensee from the requirements of 10

CFR 70.24, which requires a monitoring

system that will energize clear audible

alarms if accidental criticality occurs in

each area in which special nuclear

material is handled, used, or stored. The

proposed action would also exempt the

licensee from the requirements to

maintain emergency procedures for each

area in which this licensed special

nuclear material is handled, used, or

stored to ensure that all personnel

withdraw to an area of safety upon the

sounding of the alarm, to familiarize

personnel with the evacuation plan, and

to designate responsible individuals for

determining the cause of the alarm, and

to place radiation survey instruments in

accessible locations for use in such an

emergency.

The proposed action is in response to

the licensee’s application dated

February 4, 1997, as supplemented by

letter on March 19, 1997.

The Need for the Proposed Action

The purpose of 10 CFR 70.24 is to

ensure that if a criticality were to occur

during the handling of special nuclear

material, personnel would be alerted to

that fact and would take appropriate

action. At a commercial nuclear power

plant the inadvertent criticality with

which 10 CFR 70.24 is concerned could

occur during fuel handling operations.

The special nuclear material that could

be assembled into a critical mass at a

commercial nuclear power plant is in

the form of nuclear fuel; the quantity of

other forms of special nuclear material

that is stored on site is small enough to

preclude achieving a critical mass.

Because the fuel is not enriched beyond

5.0 weight percent Uranium-235 and

because commercial nuclear plant

licensees have procedures and features

designed to prevent inadvertent

criticality, the staff has determined that

it is unlikely that an inadvertent

criticality could occur due to the

handling of special nuclear material at

a commercial power reactor. The

requirements of 10 CFR 70.24, therefore,

are not necessary to ensure the safety of

personnel during the handling of special

nuclear materials at commercial power

reactors.

Environmental Impacts of the Proposed

Action

The Commission has completed its

evaluation of the proposed action and

concludes that there is no significant

environmental impact if the exemption

is granted. Inadvertent or accidental

criticality will be precluded through

compliance with the Catawba Nuclear

Station Technical Specifications, the

design of the fuel storage racks

providing geometric spacing of fuel

assemblies in their storage locations,

and administrative controls imposed on

fuel handling procedures. Technical

Specifications requirements specify

reactivity limits for the fuel storage

racks and minimum spacing between

the fuel assemblies in the storage racks.

Appendix A of 10 CFR part 50,

‘‘General Design Criteria for Nuclear

Power Plants,’’ Criterion 62, requires the

criticality in the fuel storage and

handling system to be prevented by

physical systems or processes,

preferably by use of geometrically safe

configurations. This is met at Catawba,

as identified in the Technical

Specification Section 3.9 and in the

Updated Final Safety Analysis Report

(UFSAR) Section 9.1, by detailed

procedures that must be available for

use by refueling personnel. Therefore, as

stated in the Technical Specifications,

these procedures, the Technical

Specifications requirements, and the

design of the fuel handling equipment

with built-in interlocks and safety

features, provide assurance that no

incident could occur during refueling

operations that would result in a hazard

to public health and safety. In addition,

the design of the facility does not

include provisions for storage of fuel in

a dry location.

UFSAR Section 9.1.1, New Fuel

Storage, states that new fuel will

normally be stored in the spent fuel

pool serving the respective unit and that

it may also be stored in the fuel transfer

canal. The fuel assemblies are stored in

five racks in a row having a nominal

center-to-center distance of 2 feet 13⁄4

inches. New fuel may also be stored in

shipping containers. (Note that in none

of these locations would criticality be

possible.)

The proposed exemption would not

result in any significant radiological

impacts. The proposed exemption

would not affect radiological plant

effluents nor cause any significant

occupational exposures since the

Technical Specifications, design

controls (including geometric spacing

and design of fuel assembly storage

spaces) and administrative controls

preclude inadvertent criticality. The

amount of radioactive waste would not

be changed by the proposed exemption.

The proposed exemption does not

result in any significant nonradiological

environmental impacts. The proposed

exemption involves 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 exemption, the staff

considered denial of the requested

exemption. Denial of the exemption

would result in no change in current

environmental impacts. The

environmental impacts of the proposed
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