Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37295

pressure must be maintained below the

PORV setpoint.

In addition, to prevent damage to

reactor coolant pump seals, the operator

must maintain a minimum 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 10 CFR Part 50,

Appendix G, safety margin adds

instrument uncertainty into 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 reactor coolant

pump seals which is significantly

restricted when physical conditions

such as PORV overshoot, RCP pump

>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 reactor coolant

pump seals during normal operation.

Using Code Case N–514 would allow

the licensee to 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 PORV

setpoint for LTOP events for Byron, 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 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.

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

March 14, 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

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 Byron 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 Byron Station, Units 1

and 2.

Agencies and Persons Consulted

In accordance with its stated policy,

on June 19, 1996, 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 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 March 14, 1996, 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 Byron Public Library

District 109 N. Franklin, P. O. Box 434,

Byron, Illinois 61010.

Dated at Rockville, Maryland, this 11th day

of July 1996.

For the Nuclear Regulatory Commission.

George F. Dick, Jr.,

Project Manager, Project Directorate III–2,

Division of Reactor Project—III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 96–18137 Filed 7–16–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the
37296 Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from June 22,

1996, through July 5, 1996. The last

biweekly notice was published on July

3, 1996 (61 FR 34884).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By August 16, 1996, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,
Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37297

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1-(800) 248-5100 (in Missouri

1-(800) 342-6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment 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 room for the particular

facility involved.

Carolina Power & Light Company, et

al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of amendments request: April 4,

1996

Description of amendments request:

The proposed amendments would

revise the Technical Specifications (TS)

to add an allowance to complete a TSrequired surveillance within 24 hours of

discovery of a missed surveillance in

accordance with the guidance of

Generic Letter (GL) 87-09, ‘‘Sections 3.0

and 4.0 of the Standard Technical

Specifications (STS) on the

Applicability of Limiting Conditions for

Operation and Surveillance

Requirements’’ and NUREG-1433,

‘‘Standard Technical Specifications,

General Electric Plants, BWR/4,’’

Revision 1, April 1995. Typographical

errors are being corrected and wording

adjustments are being incorporated for

consistency between plant TS

terminology and the associated Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

. The proposed amendments do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. The operational flexibility

resulting from the proposed revision to

Technical Specification 3.0.4 is consistent

with that allowed by the existing individual

LCO [limiting condition for operation] and

their associated ACTION requirements,

which provide an acceptable level of safety

for continued operation. A delay of up to 24

hours or the time of the surveillance interval,

whichever is less, provided by Technical

Specification 4.0.3 to complete a missed

surveillance reduces the probability of a

transient occurring when the affected system

or component is either out of service to allow

performance of the surveillance test, or there

is a lower level of confidence in the

operability because the normal surveillance

was exceeded. The revision to Technical

Specification 4.0.4 makes it clear that

Technical Specification 4.0.4 does not

prevent passage through or to

OPERATIONAL CONDITIONS as required to

comply with ACTION requirements. The

revision to the wording in Unit 2 Technical

Specification Table 3.12.1-1, Notation (h),

revisions to the Bases of the Technical

Specifications, and the elimination of

specific exemptions to Technical

Specifications 3.0.4 are administrative in

nature.

Based on the above, the proposed license

amendments do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed amendments do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed license amendments

do not introduce any new equipment nor do

they require any existing equipment or

systems to perform a different type of

function than they are presently designed to

perform. The proposed changes result in

improved Technical Specifications by

removing unnecessary restrictions on

changes in OPERATIONAL CONDITIONS

and facility operation, removing unnecessary

shutdowns caused by inadvertently

exceeding surveillance intervals, and

removing conflicts between various

Technical Specifications. The revision to the

wording in Unit 2 Technical Specification

Table 3.12.1-1, Notation (h), revisions to the

Bases of the Technical Specifications, and

the elimination of specific exemptions to

Technical Specification 3.0.4 are

administrative in nature.

Based on the above, the proposed license

amendments do not create a new or different

kind of accident from any previously

evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety. The operational flexibility

that results from the proposed revision to

Technical Specification 3.0.4 is consistent

with that allowed by the existing individual

LCO and associated ACTION requirements,

which provide an acceptable level of safety

for continued operation. Therefore, there is

no change in the margin of safety associated

with this change. A delay of up to 24 hours

or the length of the surveillance interval,

whichever is less, provided by Technical

Specification 4.0.3 to complete a missed

surveillance reduces the probability of a

transient occurring when the affected system

or component is either out of service to allow

performance of the surveillance test, or there

is a lower level of confidence in the

operability because the normal surveillance

was exceeded. In addition, the proposed

change acknowledges that the most common

outcome of the performance of a surveillance

is the successful demonstration that

acceptance criteria are met. The proposed

change provides the potential benefit of

avoiding a shutdown transient when required

equipment is still capable of performing its

function, and variables are still within limits.

The revision to Technical Specification 4.0.4

makes it clear that Technical Specification

4.0.4 does not prevent passage through or to

OPERATIONAL CONDITIONS as required to

comply with ACTION requirements. This

change is considered to be a clarification to

achieve consistency with existing Technical

Specification requirements. The revision to

the wording in Unit 2 Technical

Specification Table 3.12.1-1, Notation (h),

revisions to the Bases of the Technical

Specifications, and the elimination of

specific exemptions to Technical

Specification 3.0.4 are administrative in

nature.

The proposed changes would result in

improved Technical Specifications and

eliminate unnecessary plant challenges.

Based on the above, the proposed license

amendments do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297
37298 Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request:

December 6, 1995

Description of amendment request:

The proposed amendment would

relocate the crane operation and

movement of heavy loads requirements

and their bases from the Technical

Specifications (TS) to other plant

documents.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The change moves the requirements from

TS to other plant documents controlled

under 10 CFR 50.59 without affecting their

technical content. Since this change does not

alter the technical content of any

requirements, the operation of the facility in

accordance with the proposed change cannot

involve a significant increase in the

probability or consequences of an accident

previously evaluated, create the possibility of

a new or different kind of accident from any

previous evaluated, or involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: Mark Reinhart

Duke Power Company, Docket Nos. 50-

269, 270 and 50-287, Oconee Nuclear

Station, Units 1, 2 and 3, Oconee

County, South Carolina

Date of amendment request: June 6,

1996

Description of amendment request:

The proposed change would remove the

Engineered Safeguard (ES) signals that

presently open the outlet valves on the

Low Pressure Service Water (LPSW)

System coolers, LPSW-4 and LPSW-5,

on high reactor coolant system pressure

or high reactor building pressure. The

valves will continue to be operable from

the control room when needed. The

proposed change to Technical

Specification (TS) 4.5.1.1.2.a.(2) would

require that the refueling outage test

signal be applied to the LPSW pumps,

but no longer to LPSW-4 and LPSW-5,

and that the operability of the valves be

verified by cycling them from the

control room. A note would be added to

reflect that the refueling outage test of

LPSW-4 and LPSW-5 response to the ES

signal will continue to be verified until

the signal is removed from the ES

system for each unit during the

specified refueling outages. In addition,

TS 4.5.1.1.2.b would be clarified to

differentiate between test acceptance

criteria for automatic actuation of the

appropriate LPSW pumps and valves in

response to the ES signal, and

completion of travel of LPSW-4 and

LPSW-5 in response to manual

operation of the valves. A proposed

change to the Bases would also reflect

these changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Pursuant to 10CFR50.91, Duke Power

Company (Duke) has made the determination

that this amendment involves a No

Significant Hazards Consideration by

applying the standards established by NRC

regulations in 10CFR50.92. The following

discusses the basis for our analysis:

Will operation of the facility in accordance

with the proposed amendment:

A. Involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No. Eliminating the automatic signal that

opens Low Pressure Service Water (LPSW)

System valves, LPSW-4 and LPSW-5, upon

an Engineered Safeguards (ES) actuation does

not increase the probability of any accident

previously evaluated. The proposed change

would involve a delay in providing cooling

water to the Low Pressure Injection (LPI)

System coolers after a design basis accident.

Cooling water flow to the LPI coolers is

isolated during normal power operation.

During normal cold shutdown conditions,

cooling water flow to the LPI coolers is

normally open without relying on the ES

actuation signal. This cooling water flow is

needed to mitigate certain accidents, but a

delay in providing this cooling water flow

after a design basis accident does not

significantly increase the probability of any

accident previously evaluated.

Eliminating the ES actuation signal for

LPSW-4 and LPSW-5 will not increase the

consequences of an accident previously

evaluated. After a loss of coolant accident

(LOCA), operators will operate the

appropriate valves from the control room in

sufficient time to provide adequate cooling

water flow to maintain containment

temperature and pressure within acceptable

limits. Duke has also evaluated the delay of

LPSW cooling flow’s impact on core cooling

and concluded that there are no adverse

impacts on the capability to maintain core

cooling. Since the containment temperature

and pressure limits after a LOCA will not be

exceeded, this change will not increase any

potential off-site dose consequences after a

LOCA. Due to the time available for operator

action (approximately one hour), there is no

significant increase in operator burden

during this accident scenario.

B. Create the possibility of a new or

different kind of accident from the accidents

previously evaluated?

No. As stated above, due to the time

available for operator action (approximately

1 hour), there is no significant increase in

operator burden during this accident

scenario. Eliminating the ES signal that

automatically opens valves LPSW-4 and

LPSW-5 results in significantly lower flow

demand on the LPSW pumps. If all LPSW

pumps are successfully started, this could

result in a stronger pump causing deadhead

conditions on a weaker pump since the

pumps feed into the same piping system. To

prevent any potential adverse effects on the

LPSW pumps due to inadequate flow during

the initial stages of a LOCA, minimum flow

piping will be installed for the LPSW pumps

to provide adequate flowpaths for pump

minimum flow. Testing will be performed to

validate that the LPSW pumps can operate at

the chosen design value for pump minimum

flow. In addition, Duke conducted an

evaluation, based on manufacturer input, of

the thermal effects on the LPI coolers due to

delaying LPSW cooling flow. This evaluation

concluded that the 30 minute delay of LPSW

cooling flow has no adverse thermal effects

on the LPI coolers. Therefore, because there

is no significant increase in operator burden

and because there will be no adverse effects

on the LPSW pumps, LPI coolers, and

associated piping caused by the delayed

LPSW cooling flow, the proposed change will

not create the possibility of a new or different

kind of accident from the accidents

previously evaluated.

C. Involve a significant reduction in a

margin of safety?

No. There are no safety limits or limiting

safety system settings associated with the

LPSW System in the Oconee Nuclear Station

Technical Specifications. The proposed

change will not affect any existing safety

limits or limiting safety system settings. The

proposed change will not affect any existing

Limiting Conditions for Operation in the

Technical Specifications. The proposed

change involves an alternative method of

initiating cooling water flow to the LPI

coolers after a LOCA. This alternative

method will achieve the required results

since there will be no significant change in

the containment temperature and pressure

after a LOCA.

Duke has concluded based on the above

that there are no significant hazards

considerations involved in this amendment

request.

The NRC has reviewed the licensee’s

analysis and, based on this review, it
Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37299

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear Station

(GGNS), Unit 1, Claiborne County,

Mississippi

Date of amendment request: June 20,

1996

Description of amendment request:

The amendment would redefine the

secondary containment boundary to

allow the enclosure building to be

inoperable during the upcoming

refueling outage 8 (RFO 8) scheduled to

begin in October 1996. The amendment

would add a condition to the license

that the enclosure building may be

inoperable during core alterations and

movement of non-recently irradiated

fuel (i.e., fuel that has not occupied part

of a critical reactor core for 12 days)

during RFO 8 and the standby gas

treatment (SGT) system may be unable

to automatically start or achieve and

maintain the required vacuum, provided

the following conditions exist:

a. All dampers communicating

between the auxiliary building and the

enclosure building are closed.

b. The access door between the

auxiliary building and the enclosure

building is closed, except when the

access opening is being used for entry

and exit.

c. The SGT system is blocked from

automatic initiation.

d. SGT system is available for manual

initiation or the actions for Limiting

Condition for Operation 3.6.4.3 in the

Technical Specifications for GGNS are

complied with.

The non-recently irradiated fuel is

spent fuel that has decayed at least 12

days after the reactor was shut down for

refueling.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not

significantly increase the probability or

consequences of an accident previously

evaluated.

The equipment affected by the proposed

change is not considered an initiator to any

previously analyzed accident, therefore,

inoperability of the equipment does not

increase the probability of any previously

evaluated accident.

As described in Updated Final Safety

Analysis Report [for GGNS,] Chapter 15, the

accidents postulated to occur during core

alterations in addition to fuel handling

accidents are [the following]: inadvertent

criticality due to a control rod removal error

or continuous control rod withdrawal error

during refueling and the inadvertent loading

of a fuel assembly in an improper location.

These events are not postulated to result in

fuel cladding integrity damage. The only

accident postulated to occur during core

alterations that results in a significant

radioactive release is the fuel handling

accident. The proposed requirements in

conjunction with existing administrative

controls on light loads, bounds the

conditions of the current design basis fuel

handling accident analysis which concludes

that the radiological consequences are within

the acceptance criteria of NUREG 0800,

Section 15.7.4 and General Design Criteria

[GDC] 19 [of Appendix A to 10 CFR Part 50].

Therefore, the proposed changes do not

significantly increase consequences of any

previously evaluated accident.

Based on the above, the proposed changes

do not significantly increase the probability

or consequences of any accident previously

evaluated.

2. The proposed changes would not create

the possibility of a new or different kind of

accident from any previous analyzed.

The leaktightness of the enclosure building

does not affect the function of any plant

system other than the ability of the SGT

System to ensure the secondary containment

is at the specified pressure. The proposed

change in [the] normal SGT System

alignment[,] by defeating the automatic start

feature of the SGT System and the inability

to ensure secondary containment is at the

specified pressure[,] does not affect the

operation of any [other] plant system or

component. The SGT System is not relied

upon to provide normal or accident cooling

to plant systems or components. The

function of the enclosure building and the

SGT System is only to mitigate the release of

radioactivity to the environment in the event

of an accident.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes continue to ensure

that the radiological consequences are at or

below the current GGNS licensing limit.

Safety margins and analytical conservatisms

have been evaluated and are well

understood. Substantial margins are retained

to ensure that the analysis adequately bounds

all postulated event scenarios. The current

margin of safety is retained.

Specifically, the margin of safety for the

fuel handling accident is the difference

between the 10CFR100 [dose consequence

guidelines of 300 rem thyroid and 25 rem

whole- body] and the licensing limit defined

by NUREG-0800, Section 15.7.4. With respect

to the control room personnel doses, the

margin of safety is the difference between the

10CFR100 [guidelines] and the licensing

limit defined by 10CFR50 [10 CFR Part 50],

Appendix A, Criterion 19 (GDC 19). The

proposed applicability continues to ensure

that the whole-body and thyroid doses at the

exclusion area and low population zone

boundaries[,] as well as control room doses[,]

are at or below the corresponding licensing

limit. The margin of safety is unchanged;

therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

In excess to the margin of safety supplied

by the licensing limits of NUREG-0800 and

GDC 19, the proposed change incorporates an

additional layer of conservative

requirements. The proposed change leaves in

effect a redefined secondary containment

boundary which will provide a low leakage

boundary (consisting of the primary

containment and the auxiliary building) by

automatically isolating in the event of the

design basis fuel handling accident and

requires that the SGT System be available for

manual initiation when desired. These

requirements will ensure that doses will be

even lower than those calculated.

Therefore, the proposed changes do not

result in a significant reduction in a margin

of safety.

Based on the above evaluation, operation

in accordance with the proposed amendment

involves no significant hazards

considerations.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-424 and 50-

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of amendment request: June 17,

1996

Description of amendment request:

The proposed amendments would

revise Technical Specification Section

5.3.1 to allow use of fuel assemblies

containing fuel rods clad with ZIRLOTM.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the
37300 Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

The methodologies used in the accident

analyses remain unchanged. The proposed

change does not change or alter the design

assumptions for the systems or components

used to mitigate the consequences of an

accident. Use of ZIRLOTM fuel cladding does

not adversely affect fuel performance or

impact nuclear design methodology.

Therefore, accident analysis results are not

significantly impacted.

The operating limits will not be changed

and the analysis methods to demonstrate

operation within the limits will remain in

accordance with NRC-approved

methodologies. Other than the changes to the

fuel assemblies cladding, there are no

physical changes to the plant associated with

this Technical Specification change. A safety

analysis will continue to be performed for

each specific reload cycle to demonstrate

compliance with all fuel safety design bases.

The 10 CFR 50.46 criteria are applied to

the ZIRLOTM clad fuel rods. The use of these

fuel assemblies will not result in a change to

the reload design and safety analysis limits.

Since the original design criteria are met, the

ZIRLOTM clad fuel rods will not be an

initiator for any new accident. The clad

material is similar in chemical composition

and has similar physical and mechanical

properties as Zircaloy-4. Thus, the cladding

integrity is maintained and the structural

integrity of the fuel assembly is not affected.

ZIRLOTM cladding improves corrosion

performance and dimensional stability. Since

the dose predictions in the safety analyses

are not sensitive to the fuel rod cladding

material used, the radiological consequences

of accidents previously evaluated in the

safety analysis remain valid.

The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because:

The possibility for a new or different kind

of accident from any accident previously

evaluated is not created since the fuel

assemblies containing ZIRLOTM clad fuel

rods will satisfy the same design bases as that

currently used for Zircaloy-4 clad fuel

assemblies. All design and performance

criteria will continue to be met and no new

single failure mechanisms have been defined.

In addition, the use of ZIRLOTM fuel

assemblies does not involve any alterations

to plant equipment or procedures which

would introduce any new or unique

operational mode or accident precursor.

Therefore, the possibility for a new or

different kind of accident from any accident

previously evaluated is not created.

The proposed change does not involve a

significant reduction in a margin of safety

because:

The margin of safety is not significantly

reduced since the ZIRLOTM clad fuel

assemblies will not change the reload design

and safety analysis limits. Their use will take

into consideration the normal core operating

conditions allowed for in the Technical

Specifications. Each specific cycle’s reload

core will continue to be specifically

evaluated using NRC approved reload design

methods and approved fuel rod design

models. This will include consideration of

the core physics analysis peaking factor and

core average linear heat rate effects.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: BurkeCounty Public Library,

412 Fourth Street, Waynesboro, Georgia

30830

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308

NRC Project Director: Herbert N.

Berkow

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-424 and 50-

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of amendment request: June 17,

1996

Description of amendment request:

The proposed amendments would

clarify the requirement of Technical

Specification Surveillance Requirement

4.8.1.1.2.j(2) that requires a pressure test

of those portions of the diesel fuel-oil

system that are designed to Section III,

Subsection ND of the American Society

of Mechanical Engineers (ASME) Code.

The system pressure test would be

performed at a pressure of 110% of the

design pressure, at least once per 10

years and only on those sections of

piping that are isolable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed clarification of T/S

[Technical Specification] 4.8.1.1.2.j(2) does

not involve a significant hazards

consideration because operation of [the

Vogtle Electric Generating Plant] with this

change would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated. The configuration of

the diesel fuel-oil system as currently

installed and operated is such that a pressure

test of 110% of design pressure would be

impractical to perform. The system contains

tanks designed for atmospheric pressure and

isolation of them and their vent lines from

the specified pressure test is not practical.

The ASME Code, Section XI, provides

alternate test methods to use when storage

tanks are involved in a system pressure test.

By clarifying this T/S requirement, the

requirements set forth in ASME Section XI

can be utilized as guidance for testing

requirements to ensure the integrity of the

diesel fuel-oil system to perform its intended

safety function.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated. There are no design

changes being made that would create a new

type of accident or malfunction and the

method and manner of plant operation

remain unchanged. Using ASME Section XI

as guidance for pressure testing the isolable

sections of piping provides assurance that the

fuel oil supply system will perform its

intended function.

3. Involve a significant reduction in a

margin of safety. There are no changes being

made to the safety limits or safety system

settings that would adversely impact plant

safety. Utilizing ASME Section XI as

guidance for determining those sections of

piping that should be pressure-tested and

atmospheric-tested will ensure proper

operation of the diesel generator fuel oil

supply system.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia

30830

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308

NRC Project Director: Herbert N.

Berkow

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: April 10,

1996

Description of amendment request:

The proposed changes bring the

surveillance requirements to

conformance with Amendment No. 196

issued September 19, 1995.

Additionally, this request changes

frequency notation for a group of

surveillance requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards
Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37301

consideration (SHC), which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence of the

consequences of an accident previously

evaluated.

The proposed amendment extends the

interval between successive refueling interval

surveillances to once every 24 months for

those surveillances evaluated herein, and to

make administrative changes serving to

conform the Technical Specifications to

Amendment No. 196. Except for the

administrative changes, the proposed

surveillance interval changes do not involve

any change to the actual surveillance

requirements, nor does it involve any

change to the limits and restrictions on

plant operations. The reliability of systems

and components relied upon to prevent or

mitigate the consequences of accidents

previously evaluated is not degraded by the

proposed change to the surveillance interval.

Assurance of system and equipment

availability is maintained. This change does

not involve any change to system or

equipment configuration. Therefore, this

change does not increase the probability of

occurrence or the consequences of an

accident previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment extends the

interval between successive refueling interval

surveillances to once every 24 months for

those surveillances evaluated herein, and to

make administrative changes serving to

conform the Technical Specifications to

Amendment No. 196. Except for the

administrative changes the proposed

surveillance interval changes do not involve

any change to the limits and restrictions in

plant operation. This change does not

involve any change to system or equipment

configuration. Therefore, this change is

unrelated to the possibility of creating a new

or different kind of accident from any

previously evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendment extends the

interval between successive refueling interval

surveillances to once every 24 months for the

surveillances evaluated herein, and to make

administrative changes serving to conform

the Technical Specifications to Amendment

No. 196. Except for the administrative

changes the proposed surveillance interval

changes do not involve any change to the

actual surveillance requirements, nor does it

involve any change to the limits and

restrictions on plant operation. The

reliability of systems and components is not

degraded by the proposed change to the

surveillance interval. Assurance of system

and equipment availability is maintained.

Therefore, it is concluded that operation of

the facility in accordance with the proposed

amendment does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: May 2,

1996

Description of amendment request:

The proposed change would remove

Technical Specification Figure 5.1,

which is used in maintaining Keff

values, and substitute in its place a

defined requirement for maximum Kinfinity for any fuel placed in the

Millstone Unit 1 spent fuel pool.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Pursuant to 10CFR50.92, NNECO

[Northeast Nuclear Energy Company] has

reviewed the proposed change and concludes

that the change does not involve a significant

hazards consideration (SHC) since the

proposed change satisfies the criteria in

10CFR50.92(c). That is, the proposed change

does not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

There are no spent fuel pool accident

conditions discussed in Chapter 15 of the

FSAR [Final Safety Analysis Report]. FSAR

section 15.8 discusses a fuel handling

accident which drops a fuel assembly into

the core during refueling. Changing the

maximum allowed fuel reactivity or allowing

gaps in the Boraflex

panels will have no effect on the

probability or consequences of a fuel

assembly drop onto the core.

Therefore, based on the above, the

proposed change to the Technical

Specifications does not involve a significant

increase in the probability or consequences

of any previously analyzed accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The reduction in the allowable fuel

reactivity in the SFP [spent fuel pool] is

conservative and does not create the

possibility of a new or different type of

accident. Allowing boraflex gaps does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The margin to safety, for this proposed

technical specification change, is to maintain

the SFP Keff to be less than or equal to 0.90.

As described in the HOLTEC analysis, gaps

in the Boraflex of up to 5 inches can exist in

every boraflex panel of every rack with

Boraflex in the SFP, with Keff still less than

0.90. This is true even if all of the gaps are

uniformly lined up at the same elevation.

These calculations conservatively assumed

4% Boraflex width shrinkage as well as the

axial Boraflex gaps. Older fuel designs were

also considered to ensure that they had not

become limiting with the reduced allowable

K-infinity limit of 1.24. With no boraflex

gaps, the maximum Keff is less than .844.

With 5 inch Boraflex gaps in every panel at

the same elevation, the maximum Keff is

0.896, which is less than 0.90. NNECO has

implemented a 1 year decay time

requirement to minimize gamma irradiation

damage to the Boraflex, and will continue to

measure via ‘‘blackness testing’’ the actual

gap size to ensure the margin of safety in

maintained.

Therefore, this change has no impact on

the margin to safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141-0270

NRC Project Director: Phillip F.

McKee

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: March

29, 1996

Description of amendment request:

The proposed amendment would add

limits associated with Departure from

Nucleate Boiling (DNB) to the Indian

Point 3 (IP3) Technical Specifications.

Basis for proposed no significant

hazards consideration determination:
37302 Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Consistent with the criteria of 10 CFR

50.92, the enclosed application is judged to

involve no significant hazards based on the

following information:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously analyzed?

Response:

The proposed amendment makes no

changes to the way in which the plant is

operated and has no effect on accident

initiators associated with analyzed transients.

The probability of previously analyzed

accidents is not increased. The proposed

amendment clarifies the relationship

between measurable parameters (RCS [reactor

coolant system] temperature, pressure, and

flow rate) and the resulting heat transfer

regime in the reactor core, as characterized

by the Departure from Nucleate Boiling

(DNB) ratio. This clarification ensures that

safety analysis initial conditions regarding

heat transfer remain valid, so that the

consequences of previously analyzed

accidents are not increased. The changes

ensure that RCS pressure, temperature, and

flow are within analytical bounds. This

ensures that the plant is operated in a

manner that will not increase the

probabilities of previously analyzed

accidents nor the consequences of previously

analyzed accidents.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response:

The proposed amendment does not involve

any modifications to plant systems,

structures, or components. The proposed

change clarifies existing limits on RCS

parameters and makes no changes to plant

setpoints or operating limits. The amendment

does not involve any physical mechanism

which could contribute to a new or different

kind of accident. The changes ensure that

RCS pressure, temperature, and flow are

within analytical bounds. This ensures that

the plant is operated in a manner that will

not create the possibility of a new [or]

different kind of accident from any

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

The proposed amendment clarifies existing

limits on the measurable parameters (RCS

temperature, pressure, and flow rate) so that

the resulting DNB value is consistent with

initial condition assumptions used in

existing safety analyses. Maintaining these

limits during normal plant operation ensures

that the existing margins of safety remain

valid. The proposed amendment does not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: Jocelyn A.

Mitchell, Acting

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: June 18,

1996

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

5.2.2, ‘‘Design Pressure and

Temperature,’’ by adding design

parameters for Main Steam Line Break

(MSLB). The MSLB analysis results in a

higher containment air temperature than

the current value in TS 5.2.2.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The accidents considered for this change

are the Loss of Coolant Accident (LOCA) and

the Main Steam Line Break (MSLB). The

proposed change ensures the design limiting

containment pressure and temperature data

specified in the TS is consistent with the

[Updated Final Safety Analysis Report]

UFSAR. Since no physical changes to the

containment are being made there will be no

change in the probability of either accident

occurring.

Detailed structural analysis presented in

Supplement 1 of Licensee Event Report (LER)

272/95-016 shows that the Design Basis

LOCA combination of pressure and

temperature result in more severe loading for

the containment concrete structure and,

therefore, bounds the temperature and

pressure scenario associated with a MSLB

accident. The pressure retaining capability of

the liner is governed by the loads generated

in the MSLB. Since containment leakage is

maintained within the limits assumed in the

Accident Analysis for either scenario there is

no change in the consequences of either

accident.

Therefore the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The changes proposed affect the postaccident condition of the containment, and

have no impact on the pre-accident

condition. Since there is no physical change

proposed the containment and all systems in

the containment will continue to perform as

designed. With no physical changes being

proposed and no change to the pre-accident

condition of the containment it can be

concluded that there will be no change in the

probability of a new or different accident

being created.

Therefore the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Although calculations indicate that some

yielding of the liner plate could occur during

a MSLB, loading is transferred to the

containment concrete structure and leakage

from the containment is maintained within

the limits assumed in the Accident Analysis.

Since containment leakage is maintained

within the limits assumed in the Accident

Analysis the proposed change does not

involve a significant change the margin of

safety provided by the containment for the

MSLB.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: June 7,

1996 (TSC 95-19)

Description of amendment request:

The proposed change would revise

Section 6 of the plant Technical

Specifications to be more closely

aligned with the Revised Standard

Technical Specifications for

Westinghouse-designed nuclear plants

(NUREG-1431). Additionally, the

proposed changes would be consistent

with the guidance provided in

Administrative Letter 95-06,

‘‘Relocation of Technical Specification

Administrative Controls Related to

Quality Assurance.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the
Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37303

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA [Tennessee Valley Authority] has

concluded that operation of SQN [Sequoyah

Nuclear Plant] Units 1 and 2 in accordance

with the proposed changes to the TS

[Technical Specification] does not involve a

significant hazards consideration. TVA’s

conclusion is based on its evaluation, in

accordance with 10 CFR 50.91(a)(1), of the

three standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed TS change is administrative.

TVA has evaluated the proposed TS changes

and has determined that the proposed

changes are administrative in nature. Certain

sections are being relocated into other

licensee documents for which those

provisions are adequately controlled by

regulatory requirements. These changes do

not affect any of the design basis accidents.

They do not involve an increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

The proposed TS change is administrative.

TVA has evaluated the proposed TS changes

and has determined that the proposed

changes are administrative in nature. Certain

sections are being relocated into other

licensee documents for which those

provisions are adequately controlled by

regulatory requirements. These changes do

not affect any of the design-basis accidents.

No modifications to any plant equipment are

involved. There are no effects on system

interactions made by these changes. They do

not create the possibility of a new or different

kind of accident from an accident previously

evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed TS change is administrative.

TVA has evaluated the proposed TS changes

and has determined that the proposed

changes are administrative in nature. Certain

sections are being relocated into other

licensee documents for which those

provisions are adequately controlled by

regulatory requirements. The margin of safety

as reported in the basis for the TSs is not

reduced. The proposed change is

administrative and does not impact any

technical information contained in the bases

of the TS.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

North Atlantic Energy Service

Company, Docket No. 50-443, Seabrook

Plant Unit No. 1, Rockingham County,

New Hampshire

Date of amendment request: June 20,

1996

Description of amendment request:

The proposed amendment would

increase the allowed time for an

inoperable service water cooling tower

loop electrical supply to be the same as

the allowed outage time for an operable

service water cooling tower loop.

Date of publication of individual

notice in Federal Register: June 26, 1996

(61 FR 33142)

Expiration date of individual notice:

July 26, 1996

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, New Hampshire

Northeast Utilities Service Company,

Docket No. 50-336, Millstone Nuclear

Power Station, Unit No. 2, New London

County, Connecticut

Date of amendment request: June 3,

1996

Description of amendment request:

The proposed amendments would

provide a one-time change to Technical

Specification 3.9.1, ‘‘Refueling

Operations, Boron Concentration.’’ This

change would remove the requirement

that the boron concentration in all filled

portions of the Reactor Coolant System

be ‘‘uniform’’ and would only be

applicable during Millstone 2 Cycle 13

mid-cycle core offload.

Date of publication of individual

notice in Federal Register: June 12, 1996

(61 FR 29771)

Expiration date of individual notice:

July 12, 1996

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: May 23,

1996

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) for the

Overtemperature delta T time constants

in TS Table 2.2-1 and the Steam Line

Pressure Negative Rate High Steam Line

Isolation time constant in TS Table 3.3-

4. Date of publication of individual

notice in Federal Register: June 17, 1996

(61 FR 30639)

Expiration date of individual notice:

July 17, 1997

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: June 10,

1996

Brief description of amendment

request: The amendment proposes

changes to Technical Specification 3/

4.7.6, ‘‘Control Room Emergency Air

Conditioning System,’’ to reflect a

control room design in which the

common Salem Unit 1 and 2 control

room envelope is supplied by 2 one

hundred percent capable Control Room

Emergency Air Conditioning System

trains. Date of publication of individual

notice in Federal Register: June 24,

1996 (61 FR 32468)

Expiration date of individual notice:

July 24, 1996

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079
37304 Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices

Southern Nuclear Operating Company,

Inc., Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendment request: June 24,

1996

Description of amendment request:

The proposed amendments would

revise Technical Specification Table

4.3.1 to delete the requirement for

surveillance of the manual safety

injection to the reactor trip circuitry

until the next unit shutdown, following

which, this testing will be performed

prior to Mode 2 entry. This change is

applicable only to Unit 1, Cycle 14 and

Unit 2, Cycle 11. Date of publication of

individual notice in Federal Register:

July 3, 1996 (61 FR 34880)

Expiration date of individual notice:

August 2, 1996

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, P. O.

Box 1369, Dothan, Alabama

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items 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 rooms for the

particular facilities involved.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of application for amendments:

September 16, 1994, as supplemented

January 31, 1996.

Brief description of amendments: The

amendments revise the technical

specifications to eliminate periodic

response time testing requirements for

selected pressure and differential

pressure sensors in the reactor trip

system and engineered safety features

actuation instrumentation channels.

Date of issuance: June 26, 1996

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 84, 84, 76 and 76

Facility Operating License Nos. NPF37, NPF-66, NPF-72 and NPF-77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10393).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 26, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

November 14, 1995, as supplemented by

letters dated February 23, March 1,

March 13, March 25, March 26, May 10,

June 10, June 14, two letters dated June

25 and a letter dated June 26, 1996.

Brief description of amendments: The

proposed amendments closed out

additional open items identified in the

NRC staff’s review of the upgrade of the

Dresden and Quad Cities Technical

Specifications (TS) to the Standard

Technical Specifications (STS)

contained in NUREG-0123. The

Technical Specification Upgrade

Program (TSUP) is not a complete

adaptation of the STS. The TS upgrade

focuses on (1) integrating additional

information such as equipment

operability requirements during

shutdown conditions, (2) clarifying

requirements such as limiting

conditions for operation and action

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TS based on

the licensee’s responses to Generic

Letter (GL), and (4) relocating specific

items to more appropriate TS locations.

Date of issuance: June 28, 1996

Effective date: June 28, 1996

Amendment Nos.: 150, 145, 171, and

167

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications and operating licenses.

Date of initial notice in Federal

Register: November 29, 1995 (60 FR

61272) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

June 28, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan

Date of application for amendment:

November 22, 1995 (NRC-95-0124)

Brief description of amendment: The

amendment revises the Technical

Specifications to remove accelerated

testing frequencies and special reporting

requirements for Fermi 2 emergency

diesel generators (EDGs) in accordance

with guidance contained in Generic

Letter 94-01, dated May 31, 1994. NRC

will issue a separate safety evaluation

on extending the allowed outage time

for the EDGs at a later date.

Date of issuance: June 20, 1996

Effective date: June 20, 1996, with full

implementation within 60 days

Amendment No.: 107

Facility Operating License No. NPF43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7550) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated
Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37305

June 20, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

December 12, 1995, as supplemented by

letter dated June 10, 1996

Description of amendment request:

The amendments revise the absolute

values in the Axial Flux Difference

(AFD) Equations to reflect the proper

AFD limit reduction in the current

Technical Specifications.

Date of issuance: July 2, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 167 and 149

Facility Operating License Nos. NPF9 and NPF-17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18166)

The June 10, 1996, letter provided

clarifying information that did not

change the scope of the December 12,

1995, application and the initial

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 2, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: August

11, 1995, as supplemented by letter

dated February 12, 1996

Brief description of amendment: The

amendment reduced the minimum

reactor coolant cold leg temperature to

541 °F from 544 °F in Technical

Specification Section 3.2.6, ‘‘Reactor

Coolant Cold Leg Temperature.’’

Date of issuance: June 24, 1996

Effective date: June 24, 1996

Amendment No.: 120

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 25706)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 24, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Florida Power and Light Company,

Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

March 20, 1996, as supplemented by

letter date April 23, 1996.

Brief description of amendments:

These amendments relocate the

requirements for surveillance testing of

the water level and pressure channel

instrumentation for the reactor coolant

system accumulators. These

amendments also modify the existing

action statements of TS 3.5.1 for

accumulators to reflect the requirements

of NUREG-1431 by requiring a 72- hour

period to restore boron concentration if

it is not within the limits, and a 1-hour

period to restore any other condition

rendering the accumulators inoperable.

Date of issuance: June 24, 1996

Effective date: June 24, 1996

Amendment Nos. 185 and 179Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25707)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 24, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

February 21, 1996, as supplemented by

letters dated May 1 and June 4, 1996.

Brief description of amendments: The

amendments revise the Technical

Specifications to change the Drywell Air

Temperature Limiting Condition for

Operation (LCO) from less than or equal

to 135°F to less than or equal to 150°F.

The proposed change would provide a

margin for the primary containment

Drywell Air Temperature LCO when

prolonged summer and high river

temperatures are experienced. Also, a

strictly editorial correction to a Final

Safety Analysis Report reference would

be made.

Date of issuance: 201 and 142

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 201 and 142

Facility Operating License Nos. DPR57 and NPF-5: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18167)

The May 1 and June 4, 1996, letters

provided clarifying information that did

not change the scope of the February 21,

1996, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 27, 1996.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

Indiana Michigan Power Company,

Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

May 19, 1995, and supplemented

October 20, 1995, and April 8, 1996

(AEP:NRC:1213A)

Brief description of amendments: The

amendments modify the neutron flux

high setpoints for one or more main

steam safety valves inoperable in

response to Westinghouse Nuclear

Safety Advisory Letter 94-001. The

associated action statements are also

revised and an exemption to TS 4.0.4 is

added to support the operability

surveillance.

Date of issuance: June 28, 1996

Effective date: June 28, 1996, with full

implementation within 45 days.

Amendment Nos.: Unit 1 - 210, Unit

2 - 195

Facility Operating License Nos. DPR58 and DPR-74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65681) The April 8, 1996, submittal

provided information clarifying the

location of the TS 4.0.4 exemption

statement. This information was within

the scope of the original application and

did not alter the staff’s no significant

hazards considerations determination.

Therefore renoticing was not warranted.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 28, 1996. No

significant hazards consideration

comments received: No.
37306 Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

January 17, 1996

Brief description of amendment: The

amendment revises the Technical

Specifications (TSs) and associated

Bases by relocating certain response

time limit tables from the TSs to the

Updated Safety Analysis Report in

accordance with the guidance of NRC

Generic Letter 93-08. The relocated

tables are for instrumentation for the

Reactor Protection System, Isolation

Actuation System, Emergency Core

Cooling System, and the Recirculation

Pump Trip System.

Date of issuance: June 25, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 73

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20850) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 25, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

December 18, 1995

Brief description of amendment: The

amendment changes the Reactor Coolant

Flow - Low Flow in Technical

Specification Table 2.2-1, ‘‘Reactor

Instrumentation Protective Trip

Setpoint Limits.’’ The proposed change

increases the allowable value from

greater than or equal to 90.1% to greater

than or equal to 90.9% of the reactor

coolant flow with four pumps operating.

As an editorial change for clarification,

the word ‘‘flow’’ is added after ‘‘reactor

coolant’’ in the above sentence.

Date of issuance: July 2, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 199

Facility Operating License No. DPR65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5815) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 2, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

June 27, 1995, as supplemented July 21,

1995

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to relocate TS

requirements for the containment purge

exhaust and supply valves, and to

remove a duplicate testing requirement

for the safety injection input from

engineered safety features from the TS.

Date of issuance: June 27, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 129

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62494) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

June 27, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Pennsylvania Power and Light

Company, Docket No. 50-387,

Susquehanna Steam Electric Station,

Unit 1, Luzerne County, Pennsylvania

Date of application for amendment:

January 26, 1996

Brief description of amendment: The

amendment deletes three residual heat

removal (RHR) system relief valves from

Technical Specification (TS) Table

3.6.3-1, ‘‘Primary Containment Isolation

Valves.’’ These valves are no longer

needed to support the steam condensing

mode of RHR and are being removed

from the plant during the Unit 1 ninth

refueling outage.

Date of issuance: June 24, 1996

Effective date: As of date of issuance

to be implemented within 60 days.

Amendment No.: 157

Facility Operating License No. NPF14: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13531)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 24, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

February 29, 1996

Brief description of amendments:

These amendments relocate

Specification 3/4.9.6, ‘‘Refueling

Platform,’’ to the Susquehanna Steam

Electric Station Technical Requirements

Manual, a document which is controlled

under the requirements of 10 CFR 50.59.

Date of issuance: July 2, 1996

Effective date: July 2, 1996

Amendment Nos.: 158 and 129

Facility Operating License Nos. NPF14 and NPF-22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR 15992)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 2, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

March 12, 1996

Brief description of amendment: The

proposed changes would remove a

requirement to cross tie safety injection

accumulators.

Date of issuance: July 3, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.
Federal Register / Vol. 61, No. 138 / Wednesday, July 17, 1996 / Notices 37307

1 15 U.S.C. § 78s(b)(1) (1988).

Amendment No.: 167

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20853) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 3, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

April 24, 1996

Brief description of amendment: The

amendment proposes to relocate

Specification 3.11.B/4.11.B ‘‘Crescent

Area Ventilation’’ and associated Bases

from the TS to an Authority controlled

procedure.

Date of issuance: June 28, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 231

Facility Operating License No. DPR59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25710)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 28, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

February 6, 1996

Brief description of amendments: The

amendments change the Technical

Specifications to lower the 125 Volt

Battery Charger surveillance amperage

from at least 200 amps to at least 170

amps.

Date of issuance: June 27, 1996

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment Nos. 183 and 164

Facility Operating License Nos. DPR70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7556) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

June 27, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

June 26, 1995, as supplemented by letter

dated February 2, 1996.

Brief description of amendment: The

amendment revised the allowed outage

time for component cooling water motor

operated containment isolation valves,

moved the list of containment isolation

valves from the technical specifications

to the final safety analysis report, and

allowed containment penetration check

valves to be used as isolation devices.

Date of issuance: June 28, 1996

Effective date: June 28, 1996, to be

implemented within 30 days of the date

of issuance.

Amendment No.: 113

Facility Operating License No. NPF30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45187)

The February 2, 1996, supplemental

letter provided additional clarifying

information and did not change the

staff’s original no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 28, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Dated at Rockville, Maryland, this 10th day

of July 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation.

[Doc. 96–18007 Filed 7–16–96; 8:45 am]

BILLING CODE 7590–O1–F

[Docket No. 55–21849–OT; ASLBP No. 96–

716–01–OT]

Emerick S. McDaniel; Notice of

Reconstitution of Board

Pursuant to the authority contained in

10 CFR § 2.721, the Presiding Officer for

Emerick S. McDaniel, with the aboveidentified Docket Number, is hereby

reconstituted by appointing

Administrative Judge Peter B. Bloch as

Presiding Officer in place of Chief

Administrative Judge B. Paul Cotter, Jr.

who is unavailable to serve.

Administrative Judge Peter A. Morris

will continue to assist the Presiding

Officer in taking evidence and preparing

the record.

All correspondence, documents and

other material shall be filed with Judge

Bloch and Judge Morris in accordance

with 10 CFR § 2.701 (1980). Their

addresses are:

Administrative Judge Peter B. Bloch,

Atomic Safety and Licensing Board

Panel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555

Administrative Judge Peter A. Morris,

10825 South Glen Road, Potomac, MD

20854

Issued at Rockville, Maryland, this 11th

day of July 1996.

B. Paul Cotter, Jr.,

Chief Administrative Judge, Atomic Safety

and Licensing Board Panel.

[FR Doc. 96–18136 Filed 7–16–96; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 37420; File No. SR–MBSCC–

96–03]

Self-Regulatory Organizations; MBS

Clearing Corporation; Notice of

Proposed Rule Change Relating to

Eliminating the Monthly Audit Package

Requirements

July 11, 1996.

Pursuant to Section 19(b)(1) of the

Securities Exchange Act of 1934

(‘‘Act’’),1 notice is hereby given that on

June 18, 1996, the MBS Clearing

Corporation (‘‘MBSCC’’) filed with the

Securities and Exchange Commission

(‘‘Commission’’) the proposed rule

change as described in Items I, II, and

III below, which items have been

prepared primarily by MBSCC. The

Commission is publishing this notice to

solicit comments on the proposed rule

change from interested persons.

I. Self-Regulatory Organization’s

Statement of the Terms of Substance of

the Proposed Rule Change

MBSCC proposes to modify its rules

and procedures to eliminate the

requirement that it provide a monthly

audit package to each participant and

the requirement that such participant

review and respond to the package.