52962 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements,

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Mr. Noel F. Dudley

(telephone 301/415–6888) between 7:30

a.m. and 4:15 p.m. (EDT). Persons

planning to attend this meeting are

urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes to the agenda, etc.,

that may have occurred.

Dated: October 4, 1996.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 96–25905 Filed 10–8–96; 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

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 September

16, 1996, through September 27, 1996.

The last biweekly notice was published

on September 25, 1996 (61 FR 50338).

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 November 8, 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
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52963

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,

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.

Baltimore Gas and Electric Company,

Docket Nos. 50-317 and 50-318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

September 10, 1996

Description of amendments request:

The proposed amendment would extend

the Engineered Safety Features

Actuation System (ESFAS) automatic

actuation logic channel functional test

surveillance interval from monthly to

quarterly. The amendment request is

based on analysis documented in

Combustion Engineering Owners Group

(CEOG) Topical Reports CEN-327

(Reference a), CEN 327, Supplement 1

(Reference b), and CEN-403, Revision 1-

A, (Reference c). We have confirmed

that the information presented in CEN327 and CEN-403 is applicable to

Calvert Cliffs, and agree with the

methodology used to develop the

topical reports. In a related matter, the

licensee, also requests that the

surveillance test interval for the

containment sump isolation valves be

extended from monthly to quarterly.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The Reactor Protective System and the

Engineered Safety Features Actuation System

(ESFAS) provide the actuation signals to

safety equipment necessary to mitigate

design basis accidents and transients. The

proposed change would increase the

surveillance test interval from monthly to

quarterly for the ESFAS automatic actuation

logic channel functional tests and associated

actuation relays. The proposed change will

also extend the containment sump isolation

valve automatic opening verification

surveillance interval from monthly to

quarterly. The ESFAS instruments and

containment sump isolation valves are not

initiators in any previously evaluated

accidents. Therefore, the proposed change

does not involve an increase in the

probability of an accident previously

evaluated.

The ESFAS automatic actuation logic

circuitry and actuation relays are essentially

digital devices, which are not subject to timerelated instrument drift. Therefore, a plantspecific instrument drift analysis for these

components is not required. However, in

support of Calvert Cliffs License Amendment

Request, dated May 27, 1994, a plant-specific

setpoint drift analysis for each sensor loop

demonstrated that the observed changes in

instrument uncertainties for the extended

surveillance test interval did not exceed the

30-day setpoint assumptions. This provides

confidence that the 90-day test interval will

not impact the ability to detect and monitor

system degradation. A review of previous

containment sump isolation valve

surveillance test procedures revealed no

valve or valve operator failures. Additionally,

single failure criteria continues to be satisfied

by two redundant and independent valves on

each unit. Therefore, the proposed changes

will not change the ability of the ESFAS

instrumentation or associated engineered

safety features equipment to respond to and

mitigate the consequences of any previously

evaluated accident.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

The proposed extended surveillance test

interval for the ESFAS instruments, actuation

relays, and containment sump isolation valve

automatic opening verification does not

involve any changes in equipment or the

function of these instruments. The proposed

change does not represent a change in the

configuration or operation of the plant. The

ESFAS setpoints will not be affected since

the ESFAS automatic actuation logic

circuitry and actuation relays are not subject

to time-related instrument drift. Therefore,

the proposed change does not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The proposed change will not affect the

functions of the ESFAS instruments or

associated equipment. Topical Reports CEN327, ‘‘RPS/ESFAS Extended Test Interval

Evaluation,’’ and CEN-327, Supplement 1,

quantified the corresponding changes in core
52964 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

melt frequency for the representative fault

tree models that were developed for Calvert

Cliffs. Additionally, the ESFAS actuation

relay failure data presented in CEN-403,

Revision 1-A, ‘‘ESFAS Subgroup Relay Test

Interval Extension,’’ justifies extending the

test interval for these relays. The proposed

change has two principal effects with

opposing impacts on core melt frequency.

The first impact is a slight increase in core

melt frequency that results from the

increased possibility of an undetected

instrumentation failure due to the extended

surveillance interval. This assumed

unavailability results from less frequent

testing. The undetected ESFAS failure

represents the potential for the failure of the

appropriate engineered safety features to

actuate when required. The opposing impact

on core melt risk is the corresponding

reduction in core melt frequency that would

result due to the reduced exposure of the

plant to test-induced transients. Topical

Report CEN-327 determined that the two

changes are nearly equal, and the net result

is no distinguishable effect on plant safety.

The NRC issued a Safety Evaluation Report

which found that the above evaluations were

acceptable for justifying the extensions in the

surveillance test intervals for the ESFAS

automatic actuation logic channel functional

tests from 30 days to 90 days. In addition to

the evaluation of risk given in Topical Report

CEN-327, we have evaluated the plant

specific risk associated with these proposed

changes and concluded that changing the

surveillance intervals from monthly to

quarterly results in a net decrease in the

annual core melt frequency.

The ESFAS setpoints will not be changed

since ESFAS automatic actuation logic

circuitry and actuation relays are not subject

to time-related instrument drift. The

conclusions of the accident analyses in the

Calvert Cliffs Updated Final Safety Analysis

Report remain valid and the safety limits

continue to be met.

Extending the containment sump isolation

valve automatic opening surveillance interval

from monthly to quarterly will not

significantly reduce the margin of safety.

Both Units 1 and 2 are provided with two

containment sump isolation valves, which

satisfy single failure criteria. Historical

review of surveillance test procedures and

Nuclear Plant Reliability Data System data

revealed no failures of these valves or

associated valve operators. We have also

evaluated the plant specific risk associated

with this proposed change to the surveillance

interval and conclude that the risk is

acceptable.

Based on the generic and plant specific risk

evaluations and the demonstrated low failure

rate of the components, we conclude that

these proposed changes 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Alexander W.

Dromerick, Acting Director

Commonwealth Edison Company,

Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of amendment request: August

29, 1996, as supplemented on

September 20, 1996

Description of amendment request:

The proposed amendments would

change the Technical Specifications to

implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leakage-Test Program,’’

with certain exceptions detailed in the

licensee’s application.

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. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment cannot affect the

probability of an accident since it involves

only changes in the containment leakage rate

testing program. There is no credible

accident which can be initiated by

containment leakage rate testing.

The proposed amendment will not affect

the consequences of a[n] accident since the

allowable containment leakage rates, which

determine the offsite consequences of a[n]

accident, are unchanged. Only the frequency

of measuring the leakage rates may be

changed.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed amendment will not create

the possibility of a new or different kind of

accident since there are no changes to any

systems, structures, or components, and no

changes in the method of operation of any

system, structure, or component.

3. Involve a significant reduction in a

margin of safety.

The proposed amendment will not involve

a significant reduction in the margin of

safety. As documented in the 10 CFR 50,

Appendix J, Option B Proposed Rule and

Final Rule published in the Federal Register,

the additional industry wide risk resulting

from the proposed change is marginal and

within acceptable limits.

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

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: August

23, 1996

Description of amendment request:

The proposed amendment would

change the technical specifications to

allow fuel enrichments of up to 5.0

weight percent uranium-235.

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. Does Not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

The calculated k-effective including

uncertainties, demonstrates substantial

margin to criticality in the fuel assembly

storage locations for both normal and

accident conditions; therefore, the

probability of a previously evaluated

accident is not significantly increased. Since

a criticality accident is demonstrated to not

be feasible under the specified conditions,

the consequences of a previously evaluated

accident are not significantly increased.

Administrative controls are utilized in order

to assure that a fuel assembly is not placed

in an unanalyzed configuration. Therefore,

this change does not involve a significant

increase in the probability or consequences

of any accident previously evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

The increase in fuel enrichment could be

considered a change in plant equipment;

however, it would only affect reactivity. The

reactivity increase has been analyzed and

shown that no new or different kinds of

accidents from any previously evaluated

exist. The proposed change does not involve

the addition of any plant equipment, nor

does it modify the method of operation of

any plant equipment. Also, the proposed

change would not alter the design or

configuration of the plant beyond the

standard functional capabilities of the

equipment. Therefore, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52965

The proposed change has been analyzed to

maintain a k-effective of less than the

criticality acceptance criteria of 0.95

including uncertainties at the 95/95

probability/confidence level for all storage

configurations. Additionally, the optimum

moderation condition for the new fuel

storage racks has been analyzed and

determined to meet the acceptance criteria of

maintaining k-effective of 0.98. The use of

physical restraints for blocking the storage

locations where fuel is prohibited to be

stored in the spent fuel pools prevents

misloading of fuel into these locations. A

dropped assembly and/or the misplacement

of a fuel assembly for each storage

configuration has been analyzed. By crediting

1000 ppm boron (ANO-2 Technical

Specifications require 1600 ppm), the 95/95

k-effective is well below 0.95. Therefore, this

change does not involve a significant

reduction in the margin of safety.

Based upon the reasoning presented above

and the previous discussion of the

amendment request, Entergy Operations

proposes that the requested change does not

involve a significant hazards consideration.

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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: August

23, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

5.0, Design Features, and would for the

most part, adopt NUREG-1432, Revision

1, improved ‘‘Standard Technical

Specifications for Combustion

Engineering Plants’’ (ISTS), for this

section of the TSs.

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. Does Not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

The proposed amendment revises the

Section 5.0, Design Features, and would for

the most part, adopt NUREG-1432 Revision 1,

‘‘Standard Technical Specifications

Combustion Engineering Plants,’’ for this

section of the technical specifications. This

proposed change will also allow the

relocation of portions of the design features

section of the technical specifications to

other licensee controlled documents that are

controlled under 10 CFR 50.59. This

approach is consistent with the NRC final

policy statement and the staff’s Technical

Specification line item improvement

program. The relocation of information to

licensee controlled documents will improve

the usability and readability of technical

specifications without changing any of the

design requirements for the facility.

This amendment request does not remove

or modify any of the design requirements for

the facility or affect any accident initiators,

conditions or assumptions for any accident

previously evaluated. Therefore, this change

does not involve a significant increase in the

probability of any accident previously

evaluated.

This amendment request is administrative

in nature and does not affect any system or

component functional requirements. This

change does not affect the operation of the

plant or affect any component that is used to

mitigate the consequences of any accident.

Therefore, this change does not involve a

significant increase in the consequences of

any accident previously evaluated.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

The relocation of existing requirements

from the technical specifications to other

licensee controlled documents and the

reformatting of the design features section of

the technical specifications to the NUREG1432 format are changes that are

administrative in nature. This change does

not modify or remove any plant design

requirement. The proposed change will not

affect any plant system or structure, nor will

it affect any system functional or operability

requirements. Consequently, no new failure

modes are introduced as a result of this

change. Therefore, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed amendment request

represents a relocation of a portion of the

information previously located in the

technical specification design features

section to other licensee controlled

documents that are controlled under 10 CFR

50.59. The proposed change is administrative

in nature because the design requirements for

the facility remain the same. The proposed

change does not represent a change in the

configuration or operation of the plant.

Therefore, this change does not involve a

significant reduction in the margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve a significant hazards

consideration.

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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Niagara Mohawk Power Corporation,

Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of amendment request: July 16,

1996

Description of amendment request:

The amendment would change the

Technical Specifications to permit the

use of 10 CFR Part 50, Appendix J,

Option B, Performance-Based

Containment Leakage Rate Testing in

accordance with the implementation

guidance in NRC’s Regulatory Guide

(RG) 1.163 dated September 1995.

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 operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

NMP1 [Nine Mile Point, Unit 1] is

currently implementing Option A of

Appendix J of 10 CFR 50 for Type, A, B, and

C testing. The proposed change to the

Technical Specifications and the Bases

would implement Option B to Appendix J of

10 CFR 50 at NMP1 for Type A, B, and C

testing. Option B would allow increased

testing intervals after satisfying certain

performance based criteria.

Appendix J describes the requirements for

leakage of the primary containment and its

components penetrating the primary

containment. The leakage testing interval of

the primary containment and its components

is not a precursor or initiator to an accident.

The primary containment and its

penetrations minimizes the leakage of

radioactivity into the environment during an

accident which pressurizes the primary

containment.

Therefore, the proposed change does not

involve a significant increase in the

probability of an accident previously

evaluated.
52966 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

The proposed change to the Technical

Specifications and the Bases would replace

the detailed and prescriptive technical

requirements contained in Option A of

Appendix J with performance based

requirements in accordance with supporting

regulatory/industry documents referenced in

Option B of Appendix J. This proposed

change includes a description of the 10 CFR

50 Appendix J Testing Program Plan in

Section 6.16 of the Technical Specifications.

This program plan, with two exceptions, is

consistent with RG 1.163. Therefore, this

program plan establishes leakage-rate test

methods, procedures, acceptance criteria and

analyses which comply with Option B of

Appendix J to 10 CFR 50.

The implementation of this program

continues to provide adequate assurance that

during a DBA-LOCA [design-basis accident/

loss-of-coolant accident], the primary

containment and its components will

continue to limit leakage rates to less than

the allowable leakage rates described in the

Technical Specifications and thereby limit

leakage consistent with the assumptions of

the accident analyses. Therefore, the

increased test intervals permitted by Option

B for the primary containment and its

penetrations will continue to implement the

safety objectives underlying the requirements

of Appendix J.

As discussed below relative to the margin

of safety, the impact of the proposed change

on the consequences of a release is negligible.

The slight increase in the risk to the

population is compensated by the

corresponding risk reduction benefits

associated with the reduction in component

cycling, stress, and wear associated with

increased test intervals.

Accordingly, operation with the proposed

change to the Technical Specifications and

the Bases will not significantly increase the

consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change would implement

Option B of Appendix J to 10 CFR 50 for

Type A, B, and C testing. Option B would

allow increased testing intervals after

satisfying certain performance based criteria.

No new plant operating modes, system

operating configurations nor failure modes

are introduced by the proposed change. The

primary containment and its penetrations

will continue to perform their accident

mitigating function.

Accordingly, operation with the proposed

change will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

A regulatory impact analysis of

implementing performance-based

requirements indicates that relaxing the

frequency of Type A, B, and C testing leads

to an increase in overall risk of

approximately two percent. As indicated in

the Staff’s Regulatory Impact Analysis, this

increase is considered to be marginal to

safety.

As indicated above, increasing test

intervals can slightly increase the risk to the

population associated with the consequences

of a release; however, this is compensated by

the corresponding risk reduction benefits

associated with the reduction in component

cycling, stress, and wear associated with

increased test intervals. Therefore, when

considering the total integrated risk, the risk

associated with increased test intervals is

negligible.

The proposed change is consistent with

current plant safety analyses. In addition, the

proposed change does not require revisions

to the design of NMP1. As such, the proposed

TS changes will maintain the same level of

reliability of the equipment associated with

containment integrity, assumed to operate in

the plant safety analysis, or provide

continued assurance that specified

parameters affecting leak rate integrity, will

remain within their acceptance limits.

The as-left leakage after performing a

required leakage test continues to be less

than 0.60 La for combined Type B and C

leakage and less than or equal to 0.75 La for

Type A leakage. These as-left acceptance

criteria and the testing frequency as

established by the 10 CFR 50 Appendix J

Testing Program Plan provide assurance that

the measured leakage rate will not exceed the

maximum allowable leakage of La during

plant operation.

Visual examination of accessible interior

and exterior surfaces of the primary

containment continues to be performed prior

to initiating a Type A test. The total number

of visual examinations performed will

continue to be three times during a 10-year

period. Therefore, visual examinations of the

primary containment will continue to allow

for the timely uncovering of evidence of

structural deterioration and satisfy the

requirements of RG 1.163.

The leakage limits of LCO 3.3.3 will

continue to be met prior to reactor coolant

system temperature exceeding 215*F and

anytime Primary Containment is required.

Satisfying these leakage limits provides

assurance that the measured leakage rate will

not exceed the maximum allowable leakage

rate of La during plant operation. Therefore,

operation with the proposed change will 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 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005-3502.

NRC Project Director: Guy S. Vissing,

Acting Director

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests: May 31,

1996

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Nuclear Power Plant (DCPP),

Unit Nos. 1 and 2 to revise 23 TS

surveillance frequencies from at least

once every 18 months to at least once

per refueling interval (nominally 24

months) and to make administrative

changes for 6 other TS to maintain

consistency for TS that are not proposed

for surveillance extension. The specific

TS changes proposed include those for

2 response time tests, 3 containment

spray system tests, and 24 ventilation

system tests.

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 six administrative changes regarding

laboratory carbon testing are administrative

changes only and do not affect the

probability or consequences of accidents.

The 23 proposed TS surveillance interval

increases from 18 to 24 months do not alter

the intent or method by which the

inspections, tests, or verifications are

conducted, do not alter the way any

structure, system, or component functions,

and do not change the manner in which the

plant is operated. The surveillance,

maintenance, and operating histories indicate

that the equipment will continue to perform

satisfactorily with longer surveillance

intervals. No recurring surveillance or

maintenance problems were identified for

response time, containment spray system, or

control room ventilation system testing.

Recurring maintenance issues on the fuel

handling building and auxiliary building

ventilation systems regarding the system

control panels and certain dampers have

been addressed. These ventilation systems

are in service during all modes of operation

and experience normal wear. None of the

problems are related to refueling frequency

testing. The monthly surveillance tests

provide assurance of system operability for

the control panels. The preventative

maintenance program for the dampers is

independent of refueling shutdowns and

provides assurance that degradation

mechanisms such as corrosion and wear are

adequately addressed.
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52967

There are no known mechanisms that

would significantly degrade the performance

of the evaluated equipment during normal

plant operation. All potential time-related

degradation mechanisms have insignificant

effects in the timeframe of interest (maximum

of 30 months). Based on the past performance

of the equipment, the probability or

consequences of accidents would not be

significantly affected by the proposed

surveillance interval increases.

Therefore, the proposed changes do 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 six administrative changes regarding

laboratory carbon testing are administrative

changes only and do not affect the type of

accidents possible.

The containment spray system and control

room, auxiliary building, and fuel handling

building ventilation systems are not

associated with the initiation of any accident.

The reactor trip and engineered safety feature

actuation system response times are assumed

in the accident analysis. However, the

proposed surveillance interval increases

would not affect the type of accidents

possible.

For the 23 proposed TS changes involving

surveillance interval increases from 18- to 24-

months, the surveillance and maintenance

histories indicate that the equipment will

continue to effectively perform their

respective design functions over the longer

operating cycles. Additionally, the increased

surveillance intervals do not result in any

physical modifications, affect safety function

performance or the manner in which the

plant is operated, or alter the intent or

method by which surveillance tests are

performed. Only a few problems have been

identified and generally have not recurred.

All potential time-related degradations have

insignificant effects in the timeframe of

interest. Therefore, the proposed changes do

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.

The six administrative changes regarding

laboratory carbon testing are administrative

changes only and do not affect the margin of

safety.

For the 23 proposed TS changes involving

18- to 24-month surveillance interval

increases, evaluation of historical

surveillance and maintenance data indicates

there have been only a few problems

experienced with the evaluated equipment.

There are no indications that potential

problems would be cycle-length dependent

or that potential degradation would be

significant for the timeframe of interest;

therefore, increasing the surveillance interval

will have little, if any, impact on any margin

of safety. There is no safety analysis impact

since these changes will have no effect on

any safety limit, protection system setpoint,

or limiting condition for operation, and there

are no hardware changes that would impact

existing safety analysis acceptance criteria.

Safety margins would not be significantly

affected by the proposed surveillance interval

increases.

Therefore, the proposed changes do 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman

Rochester Gas and Electric

Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of amendment request:

September 13, 1996

Description of amendment request:

The proposed amendment changes the

Administrative Controls Section 5.6.6 of

the Ginna Station Technical

Specifications which would allow

referencing of Revision of the Ginna

Station Pressure and Temperature

Limits Report (PTLR) for the Reactor

Coolant System (RCS) pressure and

temperature (P/T) limits and low

temperature overpressure protection

(LTOP) limits.

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. Operation of Ginna Station in

accordance with the proposed changes does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed changes

only revise the reference to the PTLR in the

Administrative Controls section of technical

specifications and correct a typographical

error. The changes complete implementation

of Generic Letter 96-03 by referencing NRC

approved methodology within the

Administrative Controls. As such, these

changes are administrative in nature and do

not impact initiators or analyzed events or

assumed mitigation of accident or transient

events. Therefore, these changes do not

involve a significant increase in the

probability or consequences of an accident

previously analyzed.

2. Operation of Ginna Station in

accordance with the proposed changes does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed changes

do not involve a physical alteration of the

plant (i.e., no new or different type of

equipment will be installed) or changes in

the methods governing normal plant

operation. The proposed changes will not

impose any new or different requirements.

Thus, this change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Operation of Ginna Station in

accordance with the proposed changes does

not involve a significant reduction in a

margin of safety. The proposed changes will

not reduce a margin of plant safety because

the changes have been shown to ensure that

the P/T and LTOP limits in the revised PTLR

continue to meet all necessary requirements

for reactor vessel integrity. These changes are

administrative in nature. As such, no

question of safety is involved, and the change

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: Rochester Public Library, 115

South Avenue, Rochester, New York

14610

Attorney for licensee: Nicholas S.

Reynolds, Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005

NRC Project Director: Alexander W.

Dromerick, Acting Director

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: August

21, 1996 (TS 96-03)

Description of amendment request:

The proposed change would result in an

amendment to Licenses DPR-77 and

DPR-79 to change the Technical

Specifications (TS) for the Sequoyah

Nuclear Plant, Units 1 and 2. The

proposed change would revise TS

3.7.1.3, ‘‘Condensate Storage Tank,’’ to

include a new mode of applicability that

reads: ‘‘Mode 4 when steam generator is

relied upon for heat removal.’’ In

addition, other proposed changes to TS

3.7.1.3 and a Bases change to Bases

Sections 3/4.3.1 and 3/4.3.2, ‘‘Protective

and Engineered Safety Features (ESF)

Instrumentation,’’ are included to

provide improvements and establish

requirements that are consistent with

Westinghouse Standard Technical
52968 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

Specifications (NUREG-1431, Revision

1).

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:

Operation of Sequoyah Nuclear Plant

(SQN) in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed TS change revises SQN’s

condensate storage tank (CST) Specification

3.7.1.3 to incorporate requirements from the

Westinghouse Standard Technical

Specification (STS) contained in NUREG1431, Revision 1. The proposed change is

consistent with the STS for ensuring that

SQN’s CST remains operable in Modes 1, 2,

3, and Mode 4 when steam generator (SG) is

relied upon for heat removal. In addition, the

proposed change provides a general TS

improvement by incorporating STS

phraseology within the action requirements.

Included with this change is an increase in

the completion time for achieving hot

shutdown. The current completion time of 6

hours is increased to 12 hours. This change

allows sufficient time, while in Mode 4, to

transition from SGs to residual heat removal

entry conditions. The 12-hour completion

time is reasonable based on operating

experience to reach the required plant

condition in an orderly manner without

challenging plant systems.

[The Tennessee Valley Authority’s] TVA’s

proposed change also includes deletion of

Surveillance Requirement (SR) 4.7.1.3.2. This

SR demonstrates operability of SQN’s

essential raw cooling water (ERCW) system

every 12 hours whenever the ERCW system

is used as a supply source for the auxiliary

feedwater (AFW) system. Deletion of this SR

is consistent with STS requirements and is

justified based on: (1) current SQN TS 3.7.4

requirements ensure operability of SQN’s

ERCW in Modes 1, 2, 3, and 4, and (2) newly

proposed Action (b) requirements ensure that

SQN’s ERCW system is verified operable

every 12 hours whenever the CST is

inoperable.

The proposed changes provide TS

requirements for SQN’s CST that are

conservative with respect to assumptions

used in SQN’s accident analysis as contained

in the Final Safety Analysis Report (FSAR).

This change does not involve a physical

modification to the plant or affect any

instrumentation setpoints. Accordingly, the

proposed changes do not involve an increase

in the probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed changes provide TS

requirements for SQN’s CST that are

conservative with respect to assumptions

used in SQN’s accident analysis as contained

in the FSAR. No new event initiator has been

created, nor has any hardware been changed.

This change does not involve a physical

change to SQN’s CST or any other system.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any previously

analyzed.

3. Involve a significant reduction in a

margin of safety.

TVA’s proposed change replaces SQN’s

CST TS requirements with TS requirements

from the Westinghouse STS (NUREG-1431,

Revision 1). The proposed change to SQN’s

CST TS to add ‘‘Mode 4 when steam

generator is relied upon for heat removal,’’

provides consistency with the mode

requirements of SQN’s AFW TS and resolves

a disparity that currently exists between

these TSs. The allowed outage time for an

inoperable CST remains unchanged and is

consistent with the allowed outage time in

STS. The proposed change to delete a SR for

verifying operability of the ERCW system is

considered acceptable based on other

existing TSs that verify operability of SQN’s

ERCW system. Overall, similarity exists

between SQN’s current CST specification and

the STS version. Consequently, with the

exception of format, the TS requirements

remain essentially unchanged.

The proposed changes provide a line-item

improvement for SQN’s CST TS that are

conservative with respect to the assumptions

used in SQN’s accident analysis as contained

in the FSAR. This changes does not involve

a setpoint change or physical modification to

the plant. Accordingly, the margin of safety

has not been reduced.

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, 1101 Broad Street, Chattanooga,

Tennessee 37402

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: August

21, 1996 (TS 96-06)

Description of amendment request:

The proposed change would result in an

amendment to Licenses DPR-77 and

DPR-79 to change the Technical

Specifications for the Sequoyah Nuclear

Plant (SQN), Units 1 and 2. The

proposed change would remove

Surveillance Requirement 4.8.1.1.1.b

that verifies the ability to transfer the

unit power supply from the unit

generator supported circuit to the

preferred power circuit. The current

SQN design and operating

configurations do not require the use of

this transfer feature making this

surveillance unnecessary.

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:

Operation of Sequoyah Nuclear Plant

(SQN) in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change will delete a

surveillance associated with a feature that

does not provide a safety function based on

the current SQN design and operating

procedures. The transfer feature that is

currently verified will either be achieved by

the system alignment or covered by the

applicable TS action requirements. This

transfer feature provided automatic system

alignment to preferred offsite power circuits

for accident mitigation purposes. This

feature, or the lack of, is not considered a

source of any accident and the proposed

change to delete the associated surveillance

will not increase the possibility of an

accident previously evaluated. Safety

functions are maintained by the current

offsite circuit alignment without the transfer

feature or associated surveillance. Therefore,

the consequences of an accident can not be

increased and may be reduced by eliminating

the use of active devices to satisfy safety

functions.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed. The offsite circuit

transfer feature is not considered to be a

source of an accident and the deletion of a

surveillance to verify the operability of this

transfer will not impact this potential.

Therefore, the deletion of Surveillance

4.8.1.1.1.b will not create the possibility of a

new or different kind of accident previously

analyzed.

3. Involve a significant reduction in a

margin of safety.

The current SQN alignment satisfies all

required offsite circuit alignments necessary

to support accident mitigation functions

without the use of active devices. In addition,

any time delays associated with the transfer

actuation to realign the offsite circuits, that

is tested by the surveillance proposed to be

deleted, are eliminated by the current

alignment. Therefore, the margin of safety

associated with the affected safety function is

not reduced and may be increased by the

elimination of active devices and their

associated time delays.

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, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52969

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: August

22, 1996 (TS 96-08)

Description of amendment request:

The proposed change would result in an

amendment to Licenses DPR-77 and

DPR-79 to change the Technical

Specifications (TSs) for the Sequoyah

Nuclear Plant, Units 1 and 2. The

proposed changes would delete TS

Table 4.8-1, ‘‘Diesel Generator

Reliability,’’ and revise TS Section 3.8.1

to allow a once per 18 month, 7-day

allowed outage time (AOT) for the

emergency diesel generators (EDGs).

The first change would remove the

accelerated testing requirements for the

EDGs in accordance with Generic Letter

96-01, ‘‘Removal of Accelerated Testing

and Special Reporting Requirements for

Emergency Diesel Generators from

Technical Specifications.’’ The second

change would revise the Units 1 and 2

TS to allow a once per 18 month 7-day

AOT for planned maintenance

activities, particularly an upcoming

major 12-year overhaul of all four EDGs.

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:

Operation of Sequoyah Nuclear Plant

(SQN) in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Deletion of Table 4.8-1, in accordance with

Generic Letter (GL) 94-01, is an

administrative change that will not impact

the plant design or operation. None of the

assumptions used in evaluating the

radiological consequences of an accident are

changed. A new or altered release path is not

created. Therefore, this change does not

involve an increase in the probability of any

accident previously evaluated.

The emergency diesel generators (EDGs)

supply backup power to the essential safety

systems in the event of a loss-of-offsite

(normal) power. The EDGs cannot initiate an

accident. The requested change will not

impact the plant design or operation. The

increased out of service time does not

invalidate assumptions used in evaluating

the radiological consequences of an accident

and does not provide a new or altered release

path. Therefore, this change does not involve

an increase in the probability of any accident

previously evaluated.

An increase in the allowed outage time

(AOT) would not change the conditions,

operating configuration, or minimum amount

of operable equipment assumed in the plant

Final Safety Analysis Report for accident

mitigation. The longer AOT would provide a

longer time window for maintenance, but

would lessen the total EDG unavailability per

year. Based on the small increase in plant

risk during maintenance, and the decrease in

overall plant risk as a result of this change,

this change will not result in a significant

increase in the consequences of an accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Deletion of Table 4.8-1, in accordance with

GL 94-01, is an administrative change that

will not impact plant the plant design or

operation. Appropriate testing, in accordance

with the Maintenance Rule, will continue.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed change to extend the AOT

for the EDGs does not alter the physical

design, or configuration of the plant. The

EDG operation remains unchanged, therefore,

this change does not create the possibility of

a new or different kind of accident from any

previously analyzed.

3. Involve a significant reduction in a

margin of safety.

Deletion of Table 4.8-1, in accordance with

GL 94-01, ensures that the requirements and

provisions of 10 CFR 50.65 and the guidance

of Regulatory Guide 1.160 are met. The

program put in place by these documents

will ensure that any degradation of the EDGs

is identified and appropriate action is taken.

Therefore, this change does not involve a

significant reduction in the margin of safety.

A change to the maintenance schedule was

performed to conform with vendor

recommendations. This change in schedule

required an increase in the duration of the 18

month and longer maintenance activities.

Due to the number of shared systems, three

of the four EDGs are required to meet all of

the safety functions for each unit. However,

the TSs conservatively assume four EDGs are

necessary for unit operation; therefore, loss of

any one EDG causes entry into a LCO action

statement on both units. Performing the

required maintenance with a 72-hour AOT

will result in more EDG unavailability per

year than would be required if the AOT was

7 days. Therefore, the 7-day AOT would not

result in a significant reduction in the margin

of safety.

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, 1101 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

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: August

28, 1996 (TS 96-07)

Description of amendment request:

The proposed change would result in an

amendment to Licenses DPR-77 and

DPR-79 to change the Technical

Specifications (TS) for the Sequoyah

Nuclear Plant, Units 1 and 2. The

proposed change would revise the

setpoint tolerance for the pressurizer

safety valves (PSVs) and main steam

safety valves (MSSVs) from plus or

minus one percent to plus or minus

three percent. An analysis performed by

Framatome Technologies Incorporated

to support this change is provided in the

licensee’s submittal. These parameters

are contained in TS 3.4.2, TS 3.4.3.1,

and Table 3.7-2. Additionally, the

sentence ‘‘Following testing, lift settings

shall be within plus or minus 1%.’’

would be added to Surveillance

Requirements (SR) 4.4.2, 4.4.3.1, and

4.7.1.1.

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:

Operation of Sequoyah Nuclear Plant

(SQN) in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The evaluation contained in Enclosure 5

[Framatome Report No. 77-1257369-01,

‘‘Safety Evaluation of Safety Valve Setpoint

Tolerance Relaxation,’’ dated August 1, 1996]

discusses the consequences of this change as

it pertains to the accidents previously

analyzed. The positive increase of the

setpoint tolerance, from one percent to three

percent, of the PSV[s] and the MSSV[s] does

not challenge the design limits of the

installed systems. This conclusion is

demonstrated by means of the reanalysis of

the bounding overpressure events. The

negative tolerance for the MSSVs, from

minus one percent to minus three percent,

will result in an increase in mass discharged

through these valves. The increase was

evaluated and the analysis indicated that the

dose release remained within the limits

required by 10 CFR 100. Based on the results

of this review, there is no increase in the

probability of a previously evaluated

accident or a significant increase in the

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed TS change will increase the

setpoint tolerance for the PSVs and the

MSSVs. This change does not involve an
52970 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

equipment addition or change in the method

the plant is operated. Therefore, the

possibility of a new or different kind of

accident from any previously analyzed is not

created.

3. Involve a significant reduction in a

margin of safety.

The proposed change is a change in the lift

setpoint tolerance of the existing valves. The

analysis demonstrates that the design limits

are not exceeded nor are the dose release

limits exceeded due to this increase in

setpoint tolerance. Additionally, the valves

will be returned to the original tolerance of

plus or minus one percent. This will ensure

that the maximum margin is retained;

therefore, the margin of safety has not been

reduced by this change.

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, 1101 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

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: August 7,

1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 1.0,

‘‘Definitions,’’ by defining a refueling

interval to be [less than or equal to] 730

days; and would revise TS 3/4.0,

‘‘Applicability,’’ TS 3/4.6.2.1,

‘‘Containment Systems -

Depressurization and Cooling Systems -

Containment Spray System,’’ and TS 3/

4.6.3.1, ‘‘Containment Systems -

Containment Isolation Valves,’’ to

reflect performing surveillance tests

during a refueling interval rather than

every 18 months.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station (DBNPS), Unit No. 1, in accordance

with these changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervals from 18 to 24

months for the containment spray system

(Surveillance Requirements 4.6.2.1.b), or the

containment isolation valves (Surveillance

Requirements 4.6.3.1.2). The proposed

change to TS 1.0, adding a definition for

‘‘Refueling Interval,’’ and the associated

proposed change to TS Bases 3/4.0, are

administrative changes associated with the

24 month cycle conversion. Initiating

conditions and assumptions remain as

previously analyzed for accidents in the

DBNPS Updated Safety Analysis Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because no

potential for a significant increase in a failure

rate of an affected system or component was

identified during these reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes. The assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because no

potential for a significant increase in a failure

rate of an affected system or component was

identified during these reviews.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these revisions

do not involve any physical changes to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated. A review of

historical 18 month surveillance data and

maintenance records support an increase in

the surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because no potential for a

significant increase in a failure rate of a

system or component was identified during

these reviews. No changes are being

proposed to the type of testing currently

being performed, only to the length of the

surveillance test interval.

3. Not involve a significant reduction in a

margin of safety because a review of the

historical 18 month surveillance data and

maintenance records identified no potential

for a significant increase in a failure rate of

a system or component due to increasing the

surveillance test interval to 24 months.

Existing system and component redundancy

is not being changed by these proposed

changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences. Therefore, there are

no significant reductions 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: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service

Company, and The Cleveland Electric

Illuminating Company, Docket No. 50-346,

Davis-Besse Nuclear Power Station, Unit No.

1, Ottawa County, Ohio

Date of amendment request:

September 4, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 6.2.3,

‘‘Facility Staff Overtime,’’ by removing

specific overtime limits and working

hours and by adding procedural

controls to perform a monthly review of

overtime hours.

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:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station (DBNPS), Unit Number 1, in

accordance with these changes would:1a. Not

involve a significant increase in the

probability of an accident previously

evaluated because no change is being made

to any accident initiator. No previously

analyzed accident scenario is changed, and

initiating conditions and assumptions remain

as previously analyzed.

The proposed change to TS 6.2.3, ‘‘Facility

Staff Overtime,’’ to relocate specific overtime

limits and working hours to the DBNPS

Updated Safety Analysis Report (USAR) is

consistent with the NRC Staff’s

determination previously provided on a
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52971

generic basis in the Safety Evaluation to

License Amendment Number 127 and 116 to

the Operating Licenses (Number NPF-10 and

NPF-15), for the San Onofre Generating

Nuclear Station, Units 2 and 3, dated

February 9, 1996. The appropriate relocation

of TS requirements, such as portions of TS

6.2.3, to licensee-controlled documents is

also addressed generically in the NRC’s

‘‘Final Policy Statement on Technical

Specification Improvements for Nuclear

Power Reactors’’, dated July 23, 1993.

The relocated overtime limits and working

hours will be subject to review and

evaluation under Section 50.59, ‘‘Changes,

Tests, and Experiments’’, of Title 10 of the

Code of Federal Regulations (10 CFR) prior

to any changes being made. The other

changes to TS 6.2.3 are editorial, with an

exception being that a new requirement has

been added for plant procedures to ensure

that an individual’s overtime is reviewed

monthly by the Plant Manager or his

designee(s) to ensure excessive hours have

not been assigned.

Overtime will remain controlled by plant

administrative procedures and USAR

requirements generally following the

guidance of the NRC’s Policy Statement on

working hours contained within Generic

Letter 82-12, ‘‘Nuclear Power Plant Staff

Working Hours.’’

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed changes do

not affect accident conditions or assumptions

used in evaluating the radiological

consequences of an accident. The proposed

changes do not alter the source term,

containment isolation or allowable

radiological releases.

The proposed changes to TS 6.2.3 only

alter the administrative location of and the

regulatory controls applicable to plant staff

specific overtime limits and working hours.

Therefore, there is no significant increase in

the consequences of an accident previously

evaluated.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

changes do not change the way the plant is

operated, and no new or different failure

modes have been defined for any plant

system or component important to safety, nor

has any limiting single failure been identified

as a result of the proposed changes. No new

or different types of failures or accident

initiators are introduced by the proposed

changes.

The proposed changes to TS 6.2.3 only

alter the administrative location of and the

regulatory controls applicable to plant staff

specific overtime limits and working hours.

Therefore, there is no possibility created for

a new or different kind of accident.

3. Not involve a significant reduction in a

margin of safety because facility staff

overtime is not an input in the calculation of

any safety margin with regard to TS Safety

Limits, Limiting Safety System Settings,

other TS Limiting Conditions for Operation

or other previously defined margins for any

structure, system, or component important to

safety.

The proposed changes to TS 6.2.3 only

alter the administrative location of and the

regulatory controls applicable to plant

specific overtime limits and working hours.

Therefore, there is no 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: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service

Company, and The Cleveland Electric

Illuminating Company, Docket No. 50-346,

Davis-Besse Nuclear Power Station, Unit No.

1, Ottawa County, Ohio

Date of amendment request:

September 12, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TS) 3/4.1.3.4,

‘‘Reactivity Control Systems - Rod Drop

Time,’’ and TS 3/4.5.2, ‘‘Emergency

Core Cooling Systems - Tavg [greater

than or equal to] 280°F,’’ to change

surveillance test intervals from every 18

months to each refueling interval

(nominally 24 months). Additionally,

the proposed amendment would remove

a footnote for TS 4.5.2.b that is no

longer applicable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station (DBNPS), Unit No. 1, in accordance

with these changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervalsfrom 18 months

to 24 months for the reactivity control

systems (Surveillance Requirement 4.1.3.4.c),

or the emergency core cooling systems

(Surveillance Requirement 4.5.2.b), or the

proposed administrative change to

Surveillance Requirement 4.5.2.b to remove a

time-conditional footnote which has expired.

Initiating conditions and assumptions remain

as previously analyzed for all accidents in

the DBNPS Updated Safety Analysis Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 months to 24 months (and

up to 30 months on a non-routine basis)

because no potential for a significant increase

in a failure rate of an affected system or

component was identified during these

reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-month Fuel Cycle,’’ dated

April 2, 1991.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes. The assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because no

potential for a significant increase in a failure

rate of an affected system or component was

identified during these reviews.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these revisions

do not involve any physical changes to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated. A review of

historical 18 month surveillance data and

maintenance records support an increase in

the surveillance test intervals from 18

months to 24 months (and up to 30 months

on a non-routine basis) because no potential

for a significant increase in a failure rate of

a system or component was identified during

these reviews. No changes are being

proposed to the type of testing currently

being performed, only to the length of the

surveillance test interval.

3. Not involve a significant reduction in a

margin of safety because a review of the

historical 18 month surveillance data and

maintenance records identified no potential

for a significant increase in a failure rate of

a system or component due to increasing the

surveillance test interval to 24 months.

Existing system and component redundancy

and operation is not being changed by these

proposed changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences. Therefore, there are

no significant reductions 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
52972 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service

Company, and The Cleveland Electric

Illuminating Company, Docket No. 50-346,

Davis-Besse Nuclear Power Station, Unit No.

1, Ottawa County, Ohio

Date of amendment request:

September 17, 1996

Description of amendment request:

The proposed amendment would revise

the surveillance interval from 18

months to less than or equal to 730

days, nominally 24 months, for

Technical Specification (TS) 3/4.5.2,

‘‘Emergency Core Cooling Systems -

ECCS Subsystems - Tavg greater than or

equal to 280 degrees F;’’ TS 3/4.6.5.1,

‘‘Containment Systems - Shield

Building - Emergency Ventilation

System;’’ TS 3/4.7.6.1, ‘‘Plant Systems -

Control Room Emergency Ventilation

System;’’ TS 3/4.7.7, ‘‘Plant Systems -

Snubbers;’’ TS 3/4.9.12, ‘‘Refueling

Operations - Storage Pool Ventilation;’’

and TS Bases 3/4.7.7 - ‘‘Snubbers.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station (DBNPS), Unit No. 1 in accordance

with these changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervals from 18

months to 24 months for the trisodium

phosphate dodecahydrate (TSP) volume,

(Surveillance Requirement 4.5.2.d.4), Shield

Building and Storage Pool Emergency

Ventilation Systems, (Surveillance

Requirements 4.6.5.1.b, 4.6.5.1.d, 4.9.12.1,

and 4.9.12.2), and the Control Room

Emergency Ventilation System, (Surveillance

Requirements 4.7.6.1.c and 4.7.6.1.e) and

Snubbers (Surveillance Requirement 4.7.7.2.b

and associated Bases 3/4.7.7). Initiating

conditions and assumptions remain as

previously analyzed for all accidents in the

DBNPS Updated Safety Analysis Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because no

potential for a significant increase in a failure

rate of a system or component was identified

during these reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes. The assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these revisions

do not involve any physical changes to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated. A review of

historical 18 month surveillance data and

maintenance records support an increase in

the surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because no potential for a

significant increase in a failure rate of a

system or component was identified during

these reviews. No changes are being

proposed to the type of testing currently

being performed, only to the length of the

surveillance test interval.

3. Not involve a significant reduction in a

margin of safety because the review results of

the historical 18 month surveillance data and

maintenance records identified no potential

for a significant increase in a failure rate of

a system or component due to increasing the

surveillance test interval to 24 months.

Existing system and component redundancy

and operation is not being changed by these

proposed changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences. Therefore, there are

no significant reductions in a margin of

safety.

The NRC staff has reviewed the

licensees’ 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 Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company,

Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: August

22, 1996

Description of amendment request:

The proposed amendment would revise

the license for each unit and the bases

for Technical Specification (TS) Section

15.3.1, ‘‘Reactor Coolant System.’’ The

licensed power level would be changed

from 1518 to 1518.5 megawatts thermal

to agree with other sections of the TSs.

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. Operation of this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

There is no physical change to the facilities

as a result of the proposed license

amendment and all Limiting Conditions for

Operation, Limiting Safety System Settings

and Safety Limits specified in the Technical

Specifications remain unchanged. The

proposed change is administrative only and

restores consistency within the PBNP license

and licensing basis. Therefore, this

amendment will not cause a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed amendment has no effect on

the physical configuration of the facilities or

the manner in which they operate. The

design and design basis of the plants remains

the same. The current plant safety analysis

therefore remains complete and accurate in

addressing the design basis events and in

analyzing plant response and consequences

for the facilities. The Limiting Conditions for

Operation, Limiting Safety System Settings

and Safety Limits specified in the Technical

Specifications for the facilities are not

affected by the proposed license amendment.

The plant conditions for which the design

basis accident analysis have been performed

remain valid. Therefore, the proposed license

amendment cannot create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

Plant safety margins are established

through the Limiting Conditions for
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52973

Operation, Limiting Safety System Settings

and Safety Limits specified in the Technical

Specifications. Since there is no change to

the physical design or operation of the plant,

there is no change to any of these margins.

Thus, the proposed license amendment does

not involve a reduction in any 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

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.

Consumers Power Company, Palisades

Plant, Van Buren County, Michigan

Date of amendment request:

December 11, 1995, supplemented by

letters dated January 18, 1996, and

September 3, 1996Brief

Description of amendment request:

The proposed amendment would revise

the Palisades Technical Specifications

(TS) Administrative Controls section

(Section 6) and other TS associated with

the administrative controls section to

adopt the format of NUREG-1432,

‘‘Standard Technical Specifications,

Combustion Engineering Plants.’’ The

amendment would also revise certain

other surveillance intervals and

administrative requirements.Date of

individual notice in the Federal

Register: September 20, 1996 (61 FR

49493)

Expiration date of individual notice:

October 21, 1996

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

Consumers Power Company, Palisades

Plant, Van Buren County, Michigan

Date of amendment request: August

14, 1996 (also refer to related

application dated January 18, 1996)

Brief Description of amendment

request: The proposed amendment

would revise the Palisades Technical

Specifications (TS) to extend the

surveillance interval frequency for the

primary coolant pump (PCP) flywheels

by one operating cycle. By letter dated

January 18, 1996, the licensee

previously submitted a request to

amend the TS to delete the requirement

to perform PCP flywheel inspections.

NRC review of the original request will

not be completed in time for the

upcoming refueling outage scheduled

for November 1996; therefore, the

licensee has submitted this separate

request to extend the surveillance

frequency by one operating cycle.Date of

individual notice in the Federal

Register: September 24, 1996 (61 FR

50054

Expiration date of individual notice:

October 24, 1996

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 9,

1996, as supplemented by letter dated

August 27, 1996

Brief description of amendment

request: The amendment would revise

the Technical Specifications to allow

the surveillance of the relief mode of

operation of each of the 20 safety/relief

valves without physically lifting the

disk off the seat at power.Date of

individual notice in the Federal

Register: September 11, 1996 (61 FR

47971)

Expiration date of individual notice:

October 11, 1996

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of amendment request: August

27, 1996

Description of amendment request:

The proposed amendment would clarify

the Technical Specifications limiting

condition for operation and surveillance

requirements for the charging pumps

and high pressure safety injection

pumps when the unit is shut down

(Modes 5 and 6). Date of publication of

individual notice in Federal Register:

September 20, 1996 (61 FR 49498)

Expiration date of individual notice:

October 21, 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

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: August

27, 1996

Brief description of amendment

request: This amendment proposes to

delete License Condition 2.C.(24)(a) for

Unit 2 which required establishment by

June 3, 1981, of regularly scheduled 8-

hour shifts without reliance on routine

use of overtime. The proposed

amendment also modifies Technical

Specification 6.2.2 for both units to

incorporate limits on overtime.

Date of publication of individual

notice in Federal Register: September

12, 1996 (61 FR 48175)

Expiration date of individual notice:

October 15, 1996

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

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
52974 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

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. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

August 16, 1996

Brief description of amendments: The

amendments would defer the

implementation date as stated in

Amendment No. 150 for Dresden, Unit

2, and Amendment No. 145 for Dresden,

Unit 3, until January 15, 1997.

Date of issuance: September 26, 1996

Effective date: Immediately, to be

implemented on or before January 15,

1997.

Amendment Nos.: 151 and 146

Facility Operating License Nos. DPR19 and DPR-25: The amendments

revised the implementation date.

Date of initial notice in Federal

Register: August 22, 1996 (61 FR 43391)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 26, 1996.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Commonwealth Edison Company,

Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

April 9, 1996

Brief description of amendments: The

amendments revise the Technical

Specifications to eliminate the main

steamline radiation monitoring system

high radiation trip function for initiating

an (1) automatic reactor scram, (2)

automatic closure of the main steamline

isolation valves, and (3) automatic

closure of the reactor recirculation water

sample line isolation valves and main

steam line drain isolation valves.

Date of issuance: September 20, 1996

Effective date: Immediately, to be

implemented within 90 days.

Amendment Nos.: 115 and 100

Facility Operating License Nos. NPF11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25701)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 20,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

February 6, 1996

Brief description of amendment: The

amendment deletes the requirement to

perform alternate train testing of

redundant components when

emergency core cooling system and

containment cooling system

components are found to be inoperable

or are to be removed from service for

maintenance.

Date of issuance: September 26, 1996

Effective date: September 26, 1996

Amendment No.: 172

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28611)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 26,

1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

November 15, 1995

Brief description of amendments: The

amendments revise the Technical

Specifications to modify Section 3/4.7.5,

‘‘Standby Nuclear Service Water Pond,’’

for the Catawba Nuclear Station, Units

1 and 2, raising the minimum water

level by 1 foot (from elevation 570 to

571 feet).

Date of issuance: September 20, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 152 and 144

Facility Operating License Nos. NPF35 and NPF-52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65676) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

September 20, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Duquesne Light Company, et al., Docket

No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport,

Pennsylvania

Date of application for amendment:

April 29, 1996, as supplemented

September 12, 1996

Brief description of amendment: This

amendment revises TS 5.3.1 to allow the

use of ZIRLO as an alternate zirconiumbased fuel rod material and removes the

word ‘‘clad’’ to be consistent with the

text of the NRC’s improved Standard

Technical Specifications (NUREG-1431).

Limited substitution of fuel rods by

ZIRLO filler rods is permitted. The

proposed revision to Note 2 on TS Table

3.9-1 to specify that the maximum

burnup in the peak fuel rod in a fuel

assembly stored in Region 2 spent fuel

racks should not exceed the NRCapproved limit for WCAP-12610 was

withdrawn by letter dated September

12, 1996.

Date of issuance: September 13, 1996

Effective date: As of date of issuance

and shall be implemented within 60

days.

Amendment No: 82

Facility Operating License No. NPF73. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25703)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 13,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52975

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: May 19,

1995, as supplemented by letters dated

July 21, 1995, and June 10, September

10 and 13, 1996

Brief description of amendment: The

amendment revises the technical

specifications to permit the reactor

building personnel airlock doors to

remain open during fuel handling.

Date of issuance: September 20, 1996

Effective date: September 20, 1996

Amendment No.: 184

Facility Operating License No. DPR51. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39437)

The additional information contained in

the supplemental letters dated July 21,

1995, and June 10, September 10 and

13, 1996, were clarifying in nature and

thus, within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 20,

1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

April 4, 1995, as supplemented April 25

and August 19, 1996

Brief description of amendment: The

amendment revised operating criteria

and requirements associated with

containment personnel air locks.

Date of issuance: September 26, 1996

Effective date: September 26, 1996

Amendment No.: 175

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39438)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 26,

1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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:

December 18, 1995, as supplemented on

May 3, June 11, July 1, July 3, and

August 22, 1996.

Brief description of amendments: The

amendments increase the authorized

rated thermal power from 2200

Megawatt-thermal (MWt) to 2300 MWt.

The amendment also approves changes

to the Technical Specifications to

implement uprated power operation.

Date of issuance: September 26, 1996

Effective date: September 26, 1996, to

be implemented within 120 days

Amendment Nos. 191 and 185Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34889) The

initial Federal Register notice included

information from the licensee’s May 3

and May 11, 1996 supplemental letters.

The July 1, July 3, and August 22, 1996

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in an Environmental

Assessment dated September 12, 1996

and in a Safety Evaluation dated

September 26, 1996. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

May 15, 1996

Brief description of amendment: The

amendment revises Technical

Specification 3/4.3.2, ‘‘Isolation

Actuation Instrumentation,’’ to establish

a range of allowable values and trip

setpoints for high temperatures in the

Main Steam Line Tunnel Lead

Enclosure.

Date of issuance: September 17, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 77

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34893) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 17,

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:

January 16, 1996

Brief description of amendment: The

amendment changes the Technical

Specification (TS) Limiting Condition

for Operation Section 3.6.1 and

Surveillance Requirement Section 4.6.1,

‘‘Primary Containment,’’ and the

corresponding Bases, as well as, adds

Administrative Controls Section 6.19,

‘‘Containment Leakage Rate Testing

Program.’’ These changes will allow the

use of the performance-based

containment leakage testing

requirements described in 10 CFR Part

50, Appendix J, Option B, for Type B,

for Type A, B, and C testing.

Date of issuance: September 20, 1996

Effective date: As of the date of

issuance.

Amendment No.: 203

Facility Operating License No. DPR65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5816) as corrected on February 29, 1996

(61 FR 7825)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 20, 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

Northern States Power Company,

Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of application for amendment:

June 22, 1995, as supplemented August

10, 1995, and March 26, 1996

Brief description of amendment: The

amendment modifies the Technical

Specification requirements for

avoidance and protection from thermalhydraulic instabilities to be consistent

with the previously approved Boiling
52976 Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices

Water Reactor Owners Group long-term

solution Option I-D described in the

Licensing Topical Report, ‘‘BWR

Owners Group Long-Term Stability

Solutions Licensing Methodology

(NEDO-31960),’’ dated June 1991, and

Supplement 1 to NEDO-31960, dated

March 1992. The amendment also adds

the fuel cycle dependent stability power

and flow limits in the Core Operating

Limits Report.

Date of issuance: September 17, 1996

Effective date: September 17, 1996,

with full implementation within 60

days

Amendment No.: 97

Facility Operating License No. DPR22. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45181)

The August 10, 1995, and March 26,

1996, letters provided a nonproprietary

version of the topical report GENE-637-

043-0295 and clarifying information,

respectively. This information was

within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards consideration determination.

Therefore, renoticing was not

warranted. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 17, 1996.No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 31,

1995

Brief description of amendment: The

amendment revised the technical

specifications to require additional

restrictions on the component cooling

water system heat exchangers.

Date of issuance: September 19, 1996

Effective date: September 19, 1996

Amendment No.: 175

Facility Operating License No. DPR40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35083) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 19,

1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

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:

July 12, 1996, as supplementedAugust

19, 1996, and August 21, 1996.

Brief description of amendment: The

amendment extends the surveillance

interval for certain instruments from 18

to 24 months.

Date of issuance: September 24, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 169

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR 42282)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 24,

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.

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260, and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

May 3, 1996 (TS 352)

Brief description of amendments: The

amendments provide administrative

changes to the technical specifications.

Date of issuance: September 18, 1996

Effective Date: September 18, 1996

Amendment Nos.: 231, 246 and 206

Facility Operating License Nos. DPR33, DPR-52 and DPR-68: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR 42284)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 18,

1996.No significant hazards

consideration comments received: None

Local Public Document Room

location: Athens Public library, 405 E.

South Street, Athens, Alabama 35611

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of application for amendment:

June 28, 1996, as supplementedAugust

30, 1996.

Brief description of amendment: The

amendment revises the Technical

Specifications (TSs) to increase the

required shutdown margin. It also

revises TSs associated with this

shutdown margin increase to allow

calculational determination of the

highest worth control rod and to relax

the action requirements in the event the

required shutdown margin is not met.

The amendment also makes appropriate

editorial changes and minor editorial

corrections to the affected TSs.

Date of issuance: September 25, 1996

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 148

Facility Operating License No. DPR28. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 31, 1996 (61 FR 40031)

The August 30, 1996, letter provided

clarifying information that did not

change the scope of the application or

affect the initial determination. The

Commission’s related evaluation of the

amendment is contained in aSafety

Evaluation dated September 25,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

April 25, 1995

Brief description of amendment: The

amendment adds a reactor water

cleanup system high blowdown

containment isolation trip function and

associated limiting condition for

operation and surveillance requirements

to the Technical Specifications.

Date of issuance: September 19,

1996Effective date: September 19, 1996,

to be implemented within 30 days of

issuance.

Amendment No.: 147

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 28, 1996 (61 FR 33777)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 19,

1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352
Federal Register / Vol. 61, No. 197 / Wednesday, October 9, 1996 / Notices 52977

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

June 10, 1996, as supplemented on

August 27, and September 5, 1996.

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant Technical Specification

4.2.b, ‘‘Steam Generator Tubes,’’ and its

associated basis, by allowing the use of

Westinghouse laser-welded sleeves to

repair defective steam generator tubes.

Date of issuance: September 24, 1996

Effective date: September 24, 1996,

and is to be implemented within 30

days of the date of issuance.

Amendment No.: 127

Facility Operating License No. DPR43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34902) The

August 27, 1996, submittal increased

the TS required sample size for inservice inspection of repaired tubes in

both SGs. The September 5, 1996,

submittal incorporated the EPRI

guidelines for SG inspection scope

expansion for repaired SG tubes into the

TS. These submittals provided

clarifying information and did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

September 24, 1996. No significant

hazards consideration comments

received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

May 1, 1996 as supplemented on May

31, August 14, August 26, and

September 11, 1996.

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant Technical Specification

4.2.b, ‘‘Steam Generator Tubes,’’ its

associated bases, and Figure TS 4.2-1 by

redefining the pressure boundary for

Westinghouse mechanical hybrid

expansion joint (HEJ) steam generator

(SG) tube sleeves.

Date of issuance: September 25, 1996

Effective date: September 25, 1996,

and is to be implemented within 30

days of the date of issuance.

Amendment No.: 128

Facility Operating License No. DPR43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25715)

The May 31, August 14, August 26, and

September 11, 1996, submittals

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

September 25, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Dated at Rockville, Maryland, this 2nd day

of October 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation

[Doc. 96–25743 Filed 10–8–96; 8:45 am]

BILLING CODE 7590–01–F

Draft Regulatory Guide; Issuance,

Availability

The Nuclear Regulatory Commission

has issued for public comment a

proposed revision of a guide in its

Regulatory Guide Series. This series has

been developed to describe and make

available to the public such information

as methods acceptable to the NRC staff

for implementing specific parts of the

NRC’s regulations, techniques used by

the staff in evaluating specific problems

or postulated accidents, and data

needed by the staff in its review of

applications for permits and licenses.

The draft guide is a proposed

Revision 3 to Regulatory Guide 1.8, and

it is temporarily identified as DG–1012,

‘‘Qualification and Training of Personel

for Nuclear Power Plants.’’ The guide

will be in Division 1, ‘‘Power Reactors.’’

This regulatory guide is being revised to

provide current guidance acceptable to

the NRC staff regarding qualifications

and training for nuclear power plant

personnel. This regulatory guide would

endorse ANSI/ANS–3.1–1993,

‘‘Selection, Qualification and Training

of Personnel for Nuclear Power Plants.’’

The draft guide has not received

complete staff review and does not

represent an official NRC staff position.

Public comments are being solicited

on the guide. Comments should be

accompanied by supporting data.

Written comments may be submitted to

the Rules Review and Directives Branch,

Division of Freedom of Information and

Publications Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555.

Copies of comments received may be

examined at the NRC Public Document

Room, 2120 L Street NW., Washington,

DC. Comments will be most helpful if

received by November 15, 1996.

Although a time limit is given for

comments on this draft guide,

comments and suggestions in

connection with items for inclusion in

guides currently being developed or

improvements in all published guides

are encouraged at any time.

Comments may be submitted

electronically, in either ASCII text or

Wordperfect format (version 5.1 or

later), by calling the NRC Electronic

Bulletin Board on FedWorld. The

bulletin board may be accessed using a

personal computer, a modem, and one

of the commonly available

communications software packages, or

directly via Internet.

If using a personal computer and

modem, the NRC subsystem on

FedWorld can be accessed directly by

dialing the toll free number: 1–800–

303–9672. Communication software

parameters should be set as follows:

parity to none, data bits to 8, and stop

bits to 1 (N,8,1). Using ANSI or VT–100

terminal emulation, the NRC NUREGs

and RegGuides for Comment subsystem

can then be accessed by selecting the

‘‘Rules Menu’’ option from the ‘‘NRC

Main Menu.’’ For further information

about options available for NRC at

FedWorld, consult the ‘‘Help/

Information Center’’ from the ‘‘NRC

Main Menu.’’ Users will find the

‘‘FedWorld Online User’s Guides’’

particularly helpful. Many NRC

subsystems and data bases also have a

‘‘Help/Information Center’’ option that

is tailored to the particular subsystem.

The NRC subsystem on FedWorld can

also be accessed by a direct dial phone

number for the main FedWorld BBS,

703–321–3339, or by using Telnet via

Internet, fedworld.gov. If using 703–

321–3339 to contact FedWorld, the NRC

subsystem will be accessed from the

main FedWorld menu by selecting the

‘‘Regulatory, Government

Administration and State Systems,’’

then selecting ‘‘Regulatory Information

Mall.’’ At that point, a menu will be

displayed that has an option ‘‘U.S.

Nuclear Regulatory Commission’’ that

will take you to the NRC Online main

menu. The NRC Online area also can be

accessed directly by typing ‘‘/go nrc’’ at

a FedWorld command line. If you access

NRC from FedWorld’s main menu, you

may return to FedWorld by selecting the

‘‘Return to FedWorld’’ option from the

NRC Online Main Menu. However, if