Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64381

VIII.A and VIII.B will be reviewed and

resolved by the Director, Office for

Analysis and Evaluation of Operational

Data.

D. The NRC’s General Counsel has the

final authority to provide legal

interpretation of the Commission’s

regulations.

IX. Effective Date

This Agreement will take effect after

it has been signed by both parties.

X. Duration

A formal review, not less than 1 year

after the effective date, will be

performed by the NRC to evaluate

implementation of the Agreement and

resolve any problems identified. This

Agreement will be subject to periodic

reviews and may be amended or

modified upon written agreement by

both parties, and may be terminated

upon 30 days written notice by either

party.

XI. Separability

If any provision(s) of this Agreement,

or the application of any provision(s) to

any person or circumstances is held

invalid, the remainder of this

Agreement and the application of such

provisions to other persons or

circumstances will not be affected.

For the U.S. Nuclear Regulatory

Commission,

James M. Taylor,

Executive Director for Operations.

For the State of Louisiana.

Dated: October 31, 1996.

Gus Von Bodungen,

Assistant Secretary, Office of Air Quality and

Radiation Protection, Department of

Environmental Quality.

[FR Doc. 96–30902 Filed 12–3–96; 8:45 am]

BILLING CODE 7590–01–M

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 November 8,

1996, through November 21, 1996. The

last biweekly notice was published on

November 19, 1996.

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 January 3, 1997, 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
64382 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Carolina Power & Light Company, et

al., Docket No. 50-400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: October

31, 1996

Description of amendment request:

The proposed change would revise the

maximum allowable water temperature

as measured at the respective intake

structures from 95°F to 94°F and will

increase the minimum main reservoir

level from 205.7 feet mean sea level to

215 feet mean sea level in Technical

Specification (TS) 3/4.7.5, Ultimate Heat

Sink.

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 amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Since the proposed change does not affect

the operation of any accident initiating

systems, the probability of occurrence of an

accident previously evaluated will not

increase. Also, none of the proposed changes

will cause plant systems to operate outside

their design limits or create the likelihood of

a radioactive release. Therefore, there would

be no increase in the consequences of an

accident previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

No new component or system level

interactions will be created by the proposed

change in ultimate heat sink level and

temperature, and no design limits will be

exceeded. This change to [Technical]

Specification 3/4.7.5 is more conservative

than the current Specification limits and will

serve only to restrict operation to a higher

reservoir level and lower temperature than

was previously allowed. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

The proposed amendment will establish a

more conservative minimum main reservoir

level such that safety-related heat exchangers

served by Emergency Service Water will

continue to remove their design-basis

accident heat loads. Establishing a higher

minimum reservoir level, combined with a

more conservative reservoir temperature

assumption, will involve an increase in the

margin of safety. Also, the proposed change

in maximum reservoir temperature from 95°F

to 94°F will not result in any reduction in the

margin of safety. A maximum pre-accident

initial water temperature of 94°F is necessary

to yield a post-accident (30-day) calculated

maximum inlet temperature less than or

equal to the design basis temperature of 95°F.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64383

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart,

Acting

Duke Power Company, Docket Nos. 50-

413 and 50-414, Catawba Nuclear

Station, Units 1 and 2, York County,

South Carolina

Date of amendment request:

November 4, 1996

Description of amendment request:

The proposed amendments would

eliminate from the Technical

Specifications, Section 4.7.13.1, the

‘‘during shutdown’’ restriction

pertaining to the 18-month Standby

Shutdown System (SSS) diesel

generator inspection. Unlike Catawba

Nuclear Station, many nuclear plants do

not have an SSS facility and associated

diesel generator. The requirements in

the Technical Specifications for the SSS

diesel generator (shared between both

units) were patterned after similar

requirements for the emergency diesel

generators. The current wording

requires that both units be shut down to

perform the subject inspection.

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 standard for determining that a

Technical Specification amendment request

involves no significant hazards

considerations requires that operation of the

facility in accordance with the requested

amendment will 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 the

margin of safety.

Criterion 1

The proposed amendment seeks to change

the surveillance requirements to allow the

SSS DG [diesel generator] periodic inspection

with one or both units on line. The

surveillance can be safely completed as

proposed without affecting unit operation.

The equipment would not be removed from

service for a time that would exceed the

current Limiting Condition For Operation or

the appropriate action statement would be

entered. The probability or consequences of

any accident previously evaluated will not be

significantly increased because the removal

of the SSS DG from service can be safely

performed while one or both units are

operating.

Criterion 2

The proposed amendment change does not

change any actual surveillance requirements.

The change simply allows the 18 month SSS

DG inspection to be performed at different

unit conditions. The performance of the

surveillance with the units operating do not

require any new component configurations

that would reduce the ability of any

equipment to mitigate an accident. The

station is not degraded beyond that which

has been previously evaluated. Therefore the

proposed change does not create the

possibility of a new or different kind of

accident.

Criterion 3

The allowed outage time for the SSS DG,

as specified by the Limiting Condition For

Operation, defines the required margin of

safety for equipment operability. Removing

the SSS DG from service for periodic

inspection and returning it to service within

the allowed outage time does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the proposed

amendments involve no significant

hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Duke Power Company, Docket Nos. 50-

269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3,

Oconee County, South Carolina

Date of amendment request: October

30, 1996

Description of amendment request:

The proposed changes would (1)

completely rewrite Technical

Specification (TS) 4.4.2 to incorporate a

prestressed concrete containment

surveillance program that is consistent

with Regulatory Guide 1.35, (2) modify

TS 3.6.7 by establishing new Limiting

Conditions for Operation and required

actions related to the structural integrity

of the reactor buildings, (3) incorporate

an editorial change to TS 6.6.3 to

reference the relocated tendon

surveillance reporting requirements,

and (4) modify TS 3.6.7 Bases to

describe the Reactor Building posttensioning TS.

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) Will the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. The proposed amendment to Oconee

Technical Specifications involves the

implementation of an enhanced surveillance

program for the reactor building prestressed

concrete containment and the assurance of

appropriate station response to abnormal

degradation of the containment structure.

The proposed change will move Oconee into

a surveillance program which is consistent

with accepted industry practice and a

published NRC regulatory position. The

adoption of Regulatory Guide 1.35 as a basis

for the periodic inspection of the reactor

building prestressed concrete containment

and clearly defined station response to any

indication of structural deterioration will

assure acquisition of sufficient data to

demonstrate that structural integrity is

maintained and, if necessary, appropriate

compensatory action(s) are taken. By assuring

that any adverse trends in the behavior of the

prestressed concrete containment are

identified and acted upon in a timely

manner, this change does not increase the

probability or consequences of an accident

previously evaluated.

2) Will the change create the possibility of

a new or different kind of accident from any

previously evaluated?

No. The proposed amendment to Oconee

Technical Specifications involves the

implementation of an enhanced surveillance

program for the reactor building prestressed

concrete containment and the assurance of

appropriate station response to abnormal

degradation of the containment structure. By

adopting Regulatory Guide 1.35 as a basis for

the surveillance inspection program for the

reactor building prestressed concrete

containment and clearly defining required

station response to any indication of

structural deterioration, sufficient data will

be obtained to demonstrate that structural

integrity is being maintained and that any

adverse behavioral trends are identified and

acted upon in a timely manner. Therefore,

the proposed amendment does not create the

possibility of any type of accident: new,

different or previously evaluated.

3) Will the change involve a significant

reduction in a margin of safety?

No. Margin of safety is associated with

confidence in the ability of the fission

product barriers (i.e., fuel and fuel cladding,

Reactor Coolant System pressure boundary,

and containment structure) to limit the level

of radiation dose to the public. The proposed

Technical Specifications amendment will

move Oconee into a surveillance program

which is consistent with accepted industry

practice and a published regulatory position.

By ensuring more timely identification of,

and response to, any adverse trend in the

behavior of the reactor building prestressed

concrete containment, continued

maintenance of the structural integrity is

enhanced. Therefore, the ability of the

containment structure to perform the

intended function of protecting the public
64384 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

from radiation dose is further assured, and no

reduction in any existing margin of safety

will occur.

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: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request:

September 9, 1996

Description of amendment request:

The proposed amendment would

modify the design features section

(Section 5.0) of the Technical

Specifications (TSs) to make the design

features section consistent with the four

criteria specified in the Commission’s

Policy Statement on TSs (58 FR 39132)

and with the guidance provided in the

NRC’s Standard Technical

Specifications, Westinghouse Plants

(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:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change reduces the content

of the technical specification (TS) design

feature section consistent with the Improved

Standard Technical Specifications (ISTS) of

NUREG-1431. The information that has been

removed is also contained in the UFSAR

[Updated Final Safety Analysis Report] or

offsite dose calculation manual (ODCM);

therefore, duplication of the information is

eliminated to improve the use of the TS.

Because the information removed from the

TS is maintained in the UFSAR or ODCM

where changes are controlled in accordance

with regulatory requirements, there is no

reduction in commitment and adequate

control is provided. Elimination of

information from the design feature section

of the TS which duplicates information in

the UFSAR enhances the usability of the TS

without reducing commitments. These

changes clarify and improve the

understanding and readability of the TS.

Since the requirements remain the same,

these changes only affect the method of

presentation and would not affect possible

initiating events for accidents previously

evaluated or any system functional

requirement. Therefore, the proposed

changes would not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The relocation of existing requirements,

the elimination of requirements which

duplicate existing information, and making

administrative improvements are all changes

that are administrative in nature. The

proposed changes will not affect any plant

system or structure, not [nor] will they affect

any system functional or operability

requirements. Consequently, no new failure

modes are introduced as a result of the

proposed changes. The proposed changes are

consistent with the ISTS, for the most part,

as plant-specific information is included in

this section. Therefore, the proposed change

will not create the possibility of a new or

different type of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed changes are administrative

in nature in that no change to the design

features of the facility are being made. The

design features section is being reformatted

to be consistent, for the most part, with the

ISTS. The proposed changes do not affect the

UFSAR design bases, accident assumptions,

or technical specification bases. In addition,

the proposed changes do not affect release

limits, monitoring equipment or practices.

Therefore, 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 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: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50-458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: October

24, 1996

Description of amendment request:

The proposed amendment would revise

the technical specifications to remove

accelerated testing requirements for the

standby diesel generators. The changes

implement the provisions of Generic

Letter (GL) 94-01, ‘‘Removal of

Accelerated Testing and Special

Reporting Requirements For Diesel

Generators’’, dated May 31, 1994.

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. This request does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

This change will provide flexibility to

structure the standby diesel generator

maintenance program based on the risk

significance of the structures, systems, and

components that are within the scope of the

Maintenance Rule. The removal of the diesel

generator accelerated testing is acceptable as

the maintenance rule applies site and system

specific performance criteria to monitor

diesel generator performance. This criteria

includes a running availability and reliability

goal as well as specific goals to monitor

maintenance preventable functional failures.

The performance criteria for the diesel

generator reliability and unavailability

established by the maintenance rule and the

causal determinations and corrective actions

required for maintenance preventable

functional failures are considered to be an

acceptable method for monitoring diesel

generator performance.

The proposed change has no effect on the

probability of the initiation of an accident,

because the emergency diesel generators do

not serve as the initiator of any event.

Additionally, as diesel generator performance

will continue to be assured by the

maintenance rule, the proposed changes do

not affect the ability to mitigate the

consequences of an accident previously

evaluated. The changes do not impact the

diesel’s design sources, operating

characteristics, system functions, or system

interrelationships. The failure mechanisms

for the accidents previously analyzed are not

affected and no additional failure modes are

created that could cause an accident that has

been previously evaluated. Since the diesel

generator performance and reliability will

continue to be assured by the maintenance

rule, the proposed changes cannot involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. This request does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

This proposed change does not involve a

change to the plant design or operation. As

a result, the proposed changes does not affect

any of the parameters or conditions that

could contribute to the initiation of any

accidents. The proposed changes only affect

the methods used to monitor and assure

diesel generator performance. The

performance criteria for both the diesel

generator reliability and unavailability

established by the maintenance rule, and the
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64385

casual determinations and corrective actions

required for maintenance preventable

functional failures, is considered by GL 94-

01 to be an acceptable method for monitoring

diesel generator performance.

No [system, structure, or component] SSC,

method of operating, or system interface is

altered by this change. The changes do not

impact the diesel’s design sources, operating

characteristics, system functions, or system

interrelationships. The failure mechanisms

for the accidents are not affected, and no

additional failure modes are created. Because

the diesel generator performance and

reliability will continue to be assured by the

maintenance rule, the proposed changes

cannot create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The request does not involve a

significant reduction in a margin to safety.

The proposed changes only affect the

methods used to monitor and assure diesel

generator performance and reliability. The

performance criteria for both the diesel

generator reliability and unavailability

established by the maintenance rule, and the

casual determinations and corrective actions

required for maintenance preventable

functional failures, is considered by GL 94-

01 to be an acceptable method for monitoring

diesel generator performance. No margin to

safety as defined in the basis for any

technical specification is impacted by these

changes. This change does not impact any

uncertainty in the design, construction, or

operation of any SSC. Diesel generator

response to accident initiators is unchanged.

No SSC, method of operating, or system

interface is altered by this change. The

changes do not impact the diesel’s design

sources, operating characteristics, system

functions, or system interrelationships.

Because the diesel generator performance

and reliability will continue to be assured by

the maintenance rule, the proposed changes

cannot involve a significant reduction in the

margin to safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50-458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request:

November 6, 1996

Description of amendment request:

The proposed amendment would revise

the River Bend Station (RBS) Fire

Hazards Analysis Report and Safety

Analysis Report to allow a deviation

from 10 CFR Part 50, Appendix R,

Section III.G.2.c with respect to the

requirement for an area wide automatic

fire suppression system in Fire Area C16. The deviation would allow a 1-hour

barrier to separate redundant trains of

post fire safe shutdown equipment

within the fire area and partial sprays

on the protected train.

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 request does not involve an increase

in the probability or consequences of an

accident previously evaluated.

The event of concern is a fire in Fire Area

C-16. The low fire loading and minimal

concentration of exposed combustible

material in Fire Area C-16 would limit fire

spread. However, for this scenario, all

unprotected equipment in Fire Area C-16

will be assumed lost. Fire Area C-16 contains

cables for both Division I and Division II

components required for post fire safe

shutdown. The loss of both divisions of

cables could preclude the ability of the plant

to achieve post fire safe shutdown. Protection

of the required Division II cables in a 1-hour

fire barrier in conjunction with a partial area,

automatic suppression system installed

above and below the protected trays will

ensure that post fire safe shutdown can be

achieved.

In summary, the probability of a fire

occurring in Fire Area C-16 is not affected.

However, if a fire were to occur in Fire Area

C-16 which caused the loss of Division I

powered components, Division II powered

components, by virtue of the 1-hour fire

barrier and partial area, automatic

suppression system, would remain available.

The low fire loading and minimal

concentration of exposed combustible

material in Fire Area C-16 would limit fire

spread. The proposed fire protection scheme

provides a level of protection commensurate

with the original design. Therefore, this

request does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The request does not create the

possibility of occurrence of a new or different

kind of accident from any accident

previously evaluated.

Fire Area C-16 will be protected by a

partial area, automatic suppression system

installed above and below the protected cable

trays. Fire suppression systems are generally

used to limit fire spread, once the heat of the

fire opens thermally sensitive sprinklers. The

low fire loading and minimal concentration

of exposed combustible material in Fire Area

C-16 would aid in limiting fire spread, and

would also limit the severity of any plausible

fire. The previous analysis assumed all

Division I components and cables in the area

would be lost, and that the installed fire

barrier would adequately protect the Division

II cables routed through C-16. The required

Division II cables will be enclosed in a 1-

hour fire barrier with a partial area,

automatic suppression system. These features

provide a level of protection commensurate

with that of the previous design. In addition,

the total combustible loading in the area

results in a maximum theoretical worst case

fire duration of less than 1-hour.

In summary, if a fire were to occur in Fire

Area C-16 which caused the loss of Division

I powered components, post fire safe

shutdown could still be achieved using

Division II. Therefore, this request does not

create the possibility of occurrence of a new

or different kind of accident from any

accident previously evaluated.

3. The request does not involve a

significant reduction in a margin of safety.

In this case, the margin of safety is implicit

rather than being explicitly expressed as a

numerical value. An implicit margin of safety

involves conditions for NRC acceptance.

Since the RBS Technical Specification Bases

do not specifically address a margin of safety

for fire protection, the SAR [Safety Analysis

Report], the NRC’s Safety Evaluation Report

(SER), and appropriate other licensing basis

documents were reviewed to determine if the

proposed change would result in a reduction

in a margin of safety. As stated, in part, in

Attachment 4 to NPF-47 [Facility Operating

License; NPF-47]:

EOI [Entergy Operations, Inc.] shall

implement and maintain in effect all

provisions of the approved fire protection

program as described in the Final Safety

Analysis Report for the facility through

Amendment 22 and as approved in the SER

dated May 1984 and Supplement 3 dated

August 1985 subject to provisions 2 and 3 ....

As discussed in the Reason for Request,

SSER [Supplemental Safety Evaluation

Report] 3 dated August 1985 states, in part:

On the basis of its evaluation the staff finds

that the applicant’s fire protection program

with approved deviations is in conformance

with the guidelines of BTP CMEB [branch

technical position, Chemical Materials and

Engineering Branch] 9.5-1, [S]sections III.G,

III.J, and III.O of Appendix R to 10 CFR [Part]

50, and GDC [General Design Criteria] 3, and

is, therefore, acceptable.

Thus, the margin of safety in this case can

be defined as conformance with the specified

fire protection guidelines.

10 CFR [Part] 50, Appendix R, Section

III.G.2, requires, in part, that redundant trains

of post fire safe shutdown equipment located

in the same fire area be separated by a 1-hour

fire barrier and, in addition, that fire

detection and an automatic fire suppression

system be installed in the are under

consideration. Since Fire Area C-16 will have

a partial area, automatic suppression system,

this fire area would deviate from the

requirements of 10 CFR [Part] 50, Appendix

R, Section III.G.2.c. However, as discussed

previously, the installed partial area,

automatic suppression system, the low fire

loading and minimal amount of exposed

combustibles compensate for the lack of a

total, area wide, automatic fire suppression
64386 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

system. There is no adverse impact on the

ability to achieve and maintain post fire safe

shutdown. Therefore, this request 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: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant Units 1 and 2, St. Lucie

County, Florida

Dates of amendment requests:

October 28, 1996 (Two letters)

Description of amendment request:

The licensee proposed to change the St.

Lucie Units 1 and 2 Technical

Specifications (TS) to implement 10

CFR 50, Appendix J, Option B, for

containment leakage testing by referring

to Regulatory Guide 1.163,

‘‘Performance-Based Containment

Leakage-Test Program.’’ Changes

include relocating the details for

containment testing to the ‘‘containment

leakage rate testing program’’ and

adding the requirements of the

containment leakage rate testing

program to TS 6.8.4, which describes

facility programs. Changes are also

proposed to remove Tables 3.6-1,

‘‘Containment Leakage Paths,’’ and 3.6-

2, ‘‘Containment Isolation Valves’’ from

TS and relocate the information to plant

procedures.

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 the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated due to the following reasons:

a)These proposed changes are all

consistent with NRC requirements and

guidance for implementation of 10 CFR 50,

Appendix J, Option B, except for the use of

Bechtel Topical Report BN-TOP-1 for type A

testing. BN-TOP-1 has been previously

approved for use in accordance with 10 CFR

50 appendix J.

b) Based on industry and NRC evaluations

performed in support of developing Option

B, these changes potentially result in a minor

increase in the consequences of an accident

previously evaluated due to the increased

testing intervals. However, the proposed

changes do not result in an increase in the

core damage frequency since the containment

system is used for mitigation purposes only.

c) These changes are expected to result in

increased attention to components with poor

leakage test history as part of the

performance-based nature of Option B, such

that the marginally increased consequences

from the expanded testing intervals may be

further reduced or negated.

Therefore, these changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.(2) Operation of the facility in

accordance with the proposed amendments

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The use of the modified specifications can

not create the possibility of a new or different

kind of accident from any previously

evaluated since the proposed amendments

will not change the physical plant or the

modes of plant operation defined in the

facility operating license. No new failure

mode is introduced due to the

implementation of a performance-based

program for containment leakage rate testing,

since the proposed changes do not involve

the addition or modification of equipment,

nor do they alter the design or operation of

affected plant systems, structures, or

components.(3) Operation of the facility in

accordance with the proposed amendments

would not involve a significant reduction in

a margin of safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are basically

unchanged by the proposed amendments.

The increase in intervals between leak-test

surveillances will not significantly reduce

the margin of safety as shown by findings in

NUREG 1493, ‘‘Performance-Based

Containment Leak-Test Program’’, which was

based on implementation of the performancebased testing of Option B.

Therefore these changes do not involve a

significant reduction in the margin of

safety.The NRC staff has reviewed the

licensee’s analysis and, based on thisreview,

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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: M. S. Ross,

Attorney, Florida Power & Light, 11770

US Highway 1, North Palm Beach, Fl

33408

NRC Project Director: Frederick J.

Hebdon

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389,St.

Lucie Plant Units 1 and 2, St. Lucie

County, Florida

Date of amendment request: October

30, 1996

Description of amendment request:

The proposed amendments will revise

Technical Specification (TS) 3/4.9.10,

‘‘Refueling Operations, Water LevelReactor Vessel.’’ The Limiting Condition

for Operation (LCO) specified for the

minimum allowed refueling water level

is not altered, but the Applicability,

Action, and Surveillance Requirements

are changed to remove inconsistencies

with the definition of Core Alterations,

and to achieve consistency with the

generic Standard Technical

Specifications for Combustion

Engineering Plants (NUREG-1432). An

editorial change is proposed for TS 3/

4.9.9, ‘‘Refueling Operations,

Containment Isolation System,’’ and, for

St. Lucie Unit 1, the LCO is modified to

conform with other related refueling

specifications.

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 the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Certain evolutions performed with the UGS

[upper guide structure] in place are not Core

Alterations, and the revised LCO 3/4.9.10

will allow these activities to be performed at

water levels other than prescribed by the

existing LCO. Since these activities are

performed with the UGS in place, the

probability that a fuel handling accident

would occur is not impacted by the proposed

changes. The minimum water level required

for Core Alterations and movement of

irradiated fuel in containment is not altered

by the proposed changes, nor are any

assumptions or conditions changed that were

used as inputs to the evaluation of fuel

handling accident consequences. The

changes to Specification 3/4.9.9 are

administrative in nature and resolve an

inconsistency between the operability

requirements for the containment isolation

system and the containment penetrations that

the system would isolate at PSL1 [Plant St.

Lucie 1]. Therefore, operation of either

facility in accordance with its proposed

amendment would not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64387

kind of accident from any accident

previously evaluated.

The proposed changes are administrative

in nature, in that the changes do not involve

the addition or modification of equipment

nor do they alter the design of plant systems.

New failure modes are not introduced, and

the physical plant or the modes of plant

operation defined in the Facility License are

not altered. Therefore, operation of either

facility in accordance with its proposed

amendment would not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The safety margin associated with a fuel

handling accident is determined, in part, by

the minimum refueling water level allowed

for conducting Core Alterations and

movement of irradiated fuel in containment.

The minimum water level required by LCO

3/4.9.10, or other factors considered as inputs

to the safety analysis, is not changed by the

proposed amendments. The revised

applicability requirements for LCO 3/4.9.9 at

PSL1 will allow the containment isolation

system to be inoperable only during those

Mode 6 conditions where Core Alterations or

irradiated fuel movements within

containment are not in progress, or each

required containment penetration is

otherwise closed. Therefore, operation of

either facility in accordance with its

proposed amendment would not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on

thisreview, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendmentrequest

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: M. S. Ross,

Attorney, Florida Power & Light, 11770

US Highway 1, North Palm Beach, Fl

33408

NRC Project Director: Frederick J.

Hebdon

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: October

23, 1996, as supplemented by letter

dated November 6, 1996.

Description of amendment request:

The proposed amendment would revise

Technical Specification 3.4.6.1,

regarding reactor coolant system leakage

detection instrumentation, to adopt the

requirements found in NUREG-1431,

‘‘Standard Technical Specifications

Westinghouse Plants,’’ for the reactor

coolant system leakage detection

instrumentation.

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 involved

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change reduces the number

of containment atmospheric radioactivity

channels which must be OPERABLE when

operating in MODES 1, 2, 3, and 4 from two

to one. This change does not significantly

increase the probability or consequences of a

previously evaluated accident since the plant

will continue to have diverse and

independent means of detecting significant

changes in the amount of leakage from the

RCS [reactor coolant system]; the normal

sump level and flow monitoring system, at

least one of the two containment atmospheric

radiation monitors, and the periodic

precision RCS water inventory balance

required by Technical Specification

surveillance requirement 4.4.6.2.1.c. In

addition, STP [South Texas Project] design

includes advanced trending displays which

can assist in detecting leakage based on

changes in the volume control tank or

pressurizer level. Other instruments, which

are not listed in the Technical Specification

related to leakage, but which can provide

indication of leakage, are the containment

pressure, temperature and humidity

indicators. Good operating practice and

commercial risk associated with long term

inoperability of both monitors assures that an

inoperable containment atmospheric

radiation monitor will be promptly returned

to service.

The proposed change also revises the

limitation on continued operation with both

containment atmospheric radiation monitors

inoperable from 72 hours to 30 days. This

change is based on the continued availability

of diverse and redundant instrumentation

discussed above to detect and indicate RCS

leakage.

The Actions required as a result of this

change include analysis of a containment

atmospheric grab sample or performance of a

precision RCS water inventory balance in

accordance with surveillance requirement

4.4.6.2.1.c. The containment normal sump

level flow monitoring system will also

promptly identify changes in RCS leakage.

Other installed instrumentation, such as

containment pressure, temperature, and

humidity, will provide indications of

significant increases in leakage. Slower

increases will be detected by the daily

inventory balance or the daily grab samples

analysis, and the three day inventory

balance.

Inoperability of the on-line automatic

containment normal sump level and flow

monitoring system can be compensated for

by the performance of a daily manual

calculation, a precision RCS inventory

balance as described in surveillance

requirement 4.4.6.2.1.c, or the other available

indications of increases in leakage such as

the containment atmospheric radiation

monitoring instruments and installed

containment temperature, pressure and

humidity instrumentation. The STP control

room design also incorporates features which

allow rapid detection of unexpected changes

in the volume control tank and pressurizer

level through available instrument trend

displays. The combination of the

compensatory measures, diverse and separate

channels, and non-TS [non-technical

specification] required instrumentation

provides a sufficient level of detection to

assure prompt identification and

quantification of leakage with an inoperable

containment normal sump level and flow

monitoring system. The allowable outage

time of 30 days provides assurance the

normal containment sump level and flow

monitoring system will be returned to service

in a reasonable amount of time.

Based on the continued availability of

adequate and redundant instrumentation to

detect changes in RCS leakage rate, this

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not require the

installation of any new or different kind of

equipment. Nor does the change involve any

significant new or different MODE of

operation of the plant. The proposed change

reduces the number of required containment

atmospheric radiation monitors, and

provides a 30 day allowed outage time for

either the containment atmosphere

radioactivity monitor or the containment

normal sump level and flow monitoring

system. Therefore, this change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

In addition, as described above, the

proposed change does not significantly

reduce a margin of safety. Small changes in

RCS leak rates are typically detected over a

relatively long period of time. Diverse

instrumentation continues to be available to

plant operators which will assist in early

detection of any change. The STP design

provides additional non-Technical

Specification human factors which assist in

assuring any changes in leakage will be

quickly detected.

The proposed change extends the amount

of time that the containment atmospheric

radiation monitors may be inoperable. The

extension is based on the continued

availability of equipment which provides a

level of detection capability adequate to

detect increases in RCS leakage and which

continues to be diverse and independent.

This protection is afforded by the continued

OPERABILITY of the containment normal

sump level and flow monitoring system, the

daily performance of a precision RCS
64388 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

inventory balance as described by

surveillance requirement 4.4.6.2.1.c or the

daily analysis of containment atmospheric

grab samples, and other instrumentation such

as pressure, temperature and humidity

indicators.

The combination of the compensatory

measures, diverse and separate channels, and

non-TS required instrumentation provides a

sufficient level of detection to assure prompt

identification and quantification of leakage

with an inoperable containment normal

sump level and flow monitoring system.

Additionally, the compensatory measure of

performing either a daily manual calculation

or precision RCS inventory balance, provides

assurance that the level of safety is

maintained when the containment normal

sump level and flow monitoring system is

inoperable. The allowable outage time of 30

days provides assurance the normal

containment sump level and flow monitoring

system will be returned to service in a

reasonable amount of time.

Based on the compensatory actions and

available installed equipment, the proposed

changes do not represent 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 standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036-5869

NRC Project Director: William D.

Beckner

Northern States Power Company,

Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: August

15, 1996

Description of amendment requests:

The proposed amendments would

revise the Containment Cooling Systems

Limiting Conditions for Operation

Technical Specifications to bring them

into conformance with recently

completed system analyses by no longer

permitting both containment spray

pumps to be inoperable at the same

time.

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 amendment[s] will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Operation of the Prairie Island plant in

accordance with the proposed changes does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. None of the proposed

changes involve a physical modification to

the plant.

These changes will require operability of at

least one containment spray pump at all

times and reduces the spray additive tank

allowable outage time from 72 hours to 24

hours. Both of these changes are more

conservative and safer than currently

required in the Prairie Island Technical

Specifications. These proposed changes do

allow one containment fan cooler train out of

service for 7 days instead of 72 hours as

allowed by current Technical Specifications.

Recent plant analyses confirm that one

containment fan cooler train with one

containment spray train is sufficient to meet

the system design bases. Since the

probability of an accident occurring is low

while one containment fan cooler train is out

of service, the probability and consequences

of an accident are not significantly increased.

In total these changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because the proposed changes, in

themselves, do not introduce a new mode of

plant operation, surveillance requirement or

involve a physical modification to the plant.

The proposed changes do require more

restrictive, safer containment spray train

operability. The proposed changes also allow

one containment fan cooler train to be out of

service for 7 days instead of 72 hours as

allowed by the current Technical

Specifications. However, this change does

not create the possibility of a new kind of

accident.

The proposed changes do no alter the

design, function, or operation of any plant

components and therefore, no new accident

scenarios are created.

Therefore, the possibility of a new or

different kind of accident from any accident

previously evaluated would not be created by

these amendments.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety. This License Amendment Request

require[s] one containment spray train to be

operable at all times which is more restrictive

than current Technical Specifications and

thus the margin of safety is not reduced.

This License Amendment Request will also

allow one containment fan cooler train to be

out of service for 7 days instead of 72 hours

as allowed by the current Technical

Specifications. Since the remaining

containment cooling components can

mitigate an accident and the probability of a

design basis accident are low during this

time, this change does not significantly

reduce the plant margin of safety.

Therefore, a significant reduction in the

margin of safety would not be involved with

these amendments.

Based on the evaluation described above,

and pursuant to 10 CFR Part 50, Section

50.91, Northern States Power Company has

determined that operation [of] the Prairie

Island Nuclear Generating Plant in

accordance with the proposed license

amendment request does not involve any

significant hazards considerations as defined

by Nuclear Regulatory Commission

regulations in 10 CFR Part 50, Section 50.92.

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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: John N. Hannon

Northern States Power Company,

Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests:

September 24, 1996, as supplemented

October 17, 1996.

Description of amendment requests:

The proposed amendments would

revise the Technical Specifications (TS)

for the Prairie Island Nuclear Generating

Plant to allow use of an alternate steam

generator tube repair criteria (elevated

F-star or EF*) in the tubesheet region

when used with the repair method of

additional roll expansion. The

amendments incorporate revised

acceptance criteria for tubes with

degradation in the tubesheet region and

enable the licensee to avoid unnecessary

plugging and sleeving of steam

generator tubes.

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 amendment[s] will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The supporting technical and safety

evaluations of the subject criterion
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64389

demonstrate that the presence of the

tubesheet will enhance the tube integrity in

the region of the hardroll by precluding tube

deformation beyond its initial expanded

outside diameter. The resistance to both tube

rupture and tube collapse is strengthened by

the presence of the tubesheet in that region.

The results of hardrolling of the tube into the

tubesheet is an interference fit between the

tube and the tubesheet. Tube rupture cannot

occur because the contact between the tube

and tubesheet does not permit sufficient

movement of tube material. The radial

preload developed by the rolling process will

secure a postulated separated tube end

within the tubesheet during all plant

conditions. In a similar manner, the

tubesheet does not permit sufficient

movement of tube material to permit

buckling collapse of the tube during

postulated LOCA [loss-of-coolant accident]

loadings.

The EF* length of roll expansion is

sufficient to preclude tube pullout from tube

degradation located below the EF* distance,

regardless of the extent of the tube

degradation. The existing Technical

Specification leakage rate requirements and

accident analysis assumptions remain

unchanged in the unlikely event that

significant leakage from this region does

occur. As noted above, tube rupture and

pullout is not expected for tubes using the

EF* criterion. Any leakage out of the tube

from within the tubesheet at any elevation in

the tubesheet is fully bounded by the existing

steam generator tube rupture analysis

included in the Prairie Island Plant USAR

[updated safety analysis report]. For plants

with partial depth roll expansion like Prairie

Island, a postulated tube separation within

the tube near the top of the roll expansion

(with subsequent limited tube axial

displacement) would not be expected to

result in coolant release rates equal to those

assumed in the USAR for a steam generator

tube rupture event due to the limited gap

between the tube and tubesheet. The

proposed plugging criterion does not

adversely impact any other previously

evaluated design basis accident.

Leakage testing of roll expanded tubes

indicates that for roll lengths approximately

equal to the EF* distance, any postulated

faulted condition primary to secondary

leakage from EF* tubes would be

insignificant.

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

Implementation of the proposed EF*

criterion does not introduce any significant

changes to the plant design basis. Use of the

criterion does not provide a mechanism to

initiate an accident outside of the region of

the expanded portion of the tube. Any

hypothetical accident as a result of any tube

degradation in the expanded portion of the

tube would be bounded by the existing tube

rupture accident analysis. Tube bundle

structural integrity will be maintained. Tube

bundle leaktightness will be maintained such

that any postulated accident leakage from

EF* tubes will be negligible with regard to

offsite doses.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety.

The use of the EF* criterion has been

demonstrated to maintain the integrity of the

tube bundle commensurate with the

requirements of Reg Guide 1.121 [≥Bases for

Plugging Degraded PWR Steam Generator

Tubes≥] (intended for indications in the free

span of tubes) and the primary to secondary

pressure boundary under normal and

postulated accident conditions. Acceptable

tube degradation for the EF* criterion is any

degradation indication in the tubesheet

region, more than the EF* distance below the

bottom of the transition between the roll

expansion and the unexpanded tube. The

safety factors used in the verification of the

strength of the degraded tube are consistent

with the safety factors in the ASME

[American Society of Mechanical Engineers]

Boiler and Pressure Vessel Code used in

steam generator design. The EF* distance has

been verified by testing to be greater than the

length of roll expansion required to preclude

both tube pullout and significant leakage

during normal and postulated accident

conditions. Resistance to tube pullout is

based upon the primary to secondary

pressure differential as it acts on the surface

area of the tube, which includes the tube wall

cross-section, in addition to the inner

diameter based area of the tube. The leak

testing acceptance criteria are based on the

primary to secondary leakage limit in the

Technical Specifications and the leakage

assumptions used in the USAR accident

analyses.

Implementation of the tubesheet plugging

criterion will decrease the number of tubes

which must be taken out of service with tube

plugs or repaired with sleeves. Both plugs

and sleeves reduce the RCS (reactor coolant

system) flow margin; thus, implementation of

the EF* criterion will maintain the margin of

flow that would otherwise be reduced in the

event of increased plugging or sleeving.

Based on the above, it is concluded that the

proposed change does not result in a

significant reduction in margin with respect

to plant safety as defined in the USAR or the

Technical Specification Bases.

Based on the evaluation described above,

and pursuant to 10 CFR Part 50, Section

50.91, Northern States Power Company has

determined that operation of the Prairie

Island Nuclear Generating Plant in

accordance with the proposed license

amendment request does not involve any

significant hazards considerations as defined

by NRC regulations in 10 CFR Part 50,

Section 50.92.

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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: John N. Hannon

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: June 10,

1996, as supplemented July 25, 1996

Description of amendment request:

The proposed amendment would

change the differential temperature

Technical Specification Allowable

Values and Trip Setpoints for the

Reactor Water Cleanup penetration

room steam leak detection function.

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 [of

occurrence] [sic] or consequences of an

accident evaluated.

FSAR section 5.2.5.1.3 addresses the

ambient and differential room ventilation

temperature leakage detection. This section

states:

‘‘...switch setpoints are based on the

temperature rise resulting from a leak at

system conditions corresponding to full

reactor power.’’

NRC Safety Evaluation on the RWCU

system steam leak detection system (related

to Amendment Number 123 to License NPF14 and Amendment Number 90 to License

NPF-22) reviewed and found acceptable the

PP&L criteria for calculating the leak

detection setpoints for the RWCU system,

which include:

1. Setpoints are selected to detect and

isolate a leak that is normally less than 25

gpm and below the flow rate corresponding

for the critical crack size for the system

piping.

2. Setpoints are set high enough to avoid

inadvertent isolation caused by normal

temperature transients or abnormal transients

caused by non-leak conditions (such as loss

of ventilation).

This NRC SER also stated that a leak rate

of 25 gpm is less than those leak rates

associated with the onset of unstable pipe

ruptures. This fact is also shown in FSAR

figure 5.2-10. This value of 25 gpm

constitutes the design basis for the steam leak

detection system.

The mixing and liquid energy addition

assumption changes in the analysis do not

affect this design basis. The analysis

calculates the resulting room temperature

increase from a 25 gpm leak. In fact, the new

assumptions provide a more accurate yet

conservative prediction of room temperature

increases. Therefore, operation of the system

is improved.
64390 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

The proposed change leads to higher

calculated room temperatures to be used in

the differential temperature setpoint

calculations. The engineering study was

reviewed to determine if the higher

calculated temperatures would have a

negative impact on the High Energy Line

Break and Leak Analysis environmental

study which provides the basis for

equipment qualification.

In determining the room temperatures, the

engineering study considers ambient

temperature setpoints at which the leaks will

be isolated. The proposed action will not

change the ambient temperature setpoints,

and actuation of these instruments will

ensure that the results of the engineering

study will not be adversely affected.

Therefore, no impact on equipment

qualification is being introduced by this

change.

FSAR chapter 15 was reviewed for

potential impacts on the accident analyses.

The 25 gpm leak outside containment is not

specifically analyzed in FSAR chapter 15.

However, other conditions which result in

coolant leakage outside containment are

analyzed in section 15.6.2 (Instrument Line

Break) and 15.6.4 (Steam System Piping

Break Outside Containment). As stated in the

NRC SER, the 25 gpm RWCU leak rate is

bounded by the analysis in FSAR section

15.6.4. FSAR section 15.6.2 also states that

leak detection actuations will initiate

operator actions, a fact that is not affected by

the proposed change. Therefore, based on a

review of FSAR chapter 15 it was concluded

that no impact on the analyzed accident

scenarios is created by the proposed change.

Based on the above discussions, it is

demonstrated that the proposed change will

not adversely impact system function or

equipment. System performance will actually

be improved since the new setpoints

eliminate spurious isolations resulting from a

less accurate model. The setpoint change has

no impact on any equipment important to

safety or any accidents previously analyzed

in the FSAR. Therefore, the proposed change

does not involve a significant increase in the

probability of occurrence 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 proposed action does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. Neither the system design basis

nor the system function will be adversely

affected. System performance will be

enhanced since spurious differential

temperature actuations will be reduced as a

result of using the more accurate, yet

conservative, COTTAP model. In addition to

this, redundant temperature isolation

function will continue to be provided by the

existing high ambient temperature detectors.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed action does not involve a

significant reduction in a margin of safety.

The Technical Specification basis for the

setpoints is to detect a leak below the flow

rate corresponding to critical crack size for

the system piping. As stated previously, the

25 gpm flow rate is an acceptable flow rate

and is used to calculate the new

temperatures.

Although the newly calculated RWCU

penetration room temperatures will be higher

(due to the improved model), the isolation

actuation will be initiated by the high

ambient temperature detectors before the

penetration room temperatures reach the

newly calculated values, as would happen

under the old model. Therefore, system

response is not adversely affected.

The current temperature values lead to

differential temperature setpoints which are

too low, causing spurious isolations. The use

of the new temperature values will reduce

the number of spurious isolations, reducing

unnecessary challenges to safety systems

during normal plant operations.

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: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request:

September 18, 1995

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS Table

4.3.1.1-1, ‘‘Reactor Protection System

Instrumentation Surveillance

Requirements’’ to reflect the change in

the calibration frequency for the Local

Power Range Monitor (LPRM) signal

from every 1000 Effective Full Power

Hours (EFPH) to every 2000 Megawatt

Days per Standard Ton (MWD/ST).

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 Technical Specifications

(TS) change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The change in the calibration frequency of

the Local Power Range Monitor (LPRM)

signal does not make any physical change to

the fuel or the manner in which the fuel

responds to a transient or accident. The

proposed TS change does not affect the

fundamental method by which the LPRMs

are calibrated. Also, the LPRM calibration

frequency is not considered an initiator of

any events analyzed in the SAR. Therefore,

calibrating the LPRMs on a different

frequency will not increase the probability of

occurrence of an accident previously

evaluated in the SAR.

The resulting nodal power uncertainty

does not exceed the nodal power uncertainty

accounted for in the existing Minimum

Critical Power Ratio (MCPR) Safety Limit;

thus, the MCPR Safety Limit is not affected

by this TS Change, and, therefore, the initial

conditions of any accident are unchanged.

Since the calibration frequency change will

not affect the course of any evaluated

accident, the consequences of an accident

previously evaluated in the SAR will not be

increased.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The change in the calibration frequency of

the Local Power Range Monitor (LPRM)

signal does not make any physical change to

the plant or the manner in which the

equipment responds to a transient or

accident. The proposed TS change does not

introduce a new mode of plant operation and

does not involve the installation of any new

equipment or instrumentation. The fuel will

continue to be operated to the same safety

limits since the Minimum Critical Power

Ratio (MCPR) Safety Limit remains

unchanged due to this TS change.

Therefore, the proposed TS change does

not create the possibility of a new or different

kind of accident, from any accident

previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The following TS Bases were reviewed for

potential reduction in the margin of safety:

2.0 Safety Limits and Limiting Safety

System Settings;

3/4.1 Reactivity Control Systems;

3/4.2.1 Average Planar Linear Heat

Generation Rate;

3/4.2.3 Minimum Critical Power Ratio:

3/4.2.4 Linear Heat Generation Rate;

3/4.3.1 Reactor Protection System

Instrumentation;

3/4.3.6 Control Rod Block Instrumentation;

3/4.3.7.7 Traversing In-Core Probe System;

The GE Thermal Analysis Basis (GETAB)

determination of the Minimum Critical

Power Ratio (MCPR) Safety Limit allows a

maximum total nodal uncertainty of the

Traversing In-Core Probe (TIP) readings of

which the Local Power Range Monitor

(LPRM).

Update uncertainty is a part. The change in

LPRM calibration frequency results in an

LPRM Update uncertainty which, when

combined with the other uncertainties which

comprise the total TIP readings uncertainty,

yields a total TIP readings nodal power

uncertainty of less than the allowed GETAB

uncertainty. Thus the change in LPRM
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64391

calibration frequency will not affect the

MCPR Safety Limit.

The LPRMs are utilized as input to the

Average Power Range Monitor (APRM) and

Rod Block Monitor (RBM) systems. The

primary safety function of the APRM system

is to initiate a scram during core-wide

neutron flux transients before the actual corewide neutron flux level exceeds the safety

analysis design basis. This prevents fuel

damage from single operator errors or

equipment malfunctions. The APRMs are

calibrated at least once per week to the plant

heat balance, utilize a radially and axially

diverse group of LPRMs as input and are

utilized to detect changes in average, not

local, power changes. Therefore, the effects of

changing the LPRM calibration frequency on

the APRM system responses will be minimal

due to any individual LPRM drift being

practically canceled out (due to diversity of

input) and/or due to the frequent

recalibration of the APRMs to an

independent power calculation (the heat

balance). Thus, changing the LPRM

calibration frequency will not impact the

capability of the APRM system to perform the

scram function, and there is no impact on

transient delta-CPRs.

The RBM system is utilized in the

mitigation of a Rod Withdrawal Error (RWE)

event. The RBM system is designed to

prevent the operator from increasing the local

power significantly when withdrawing a

control rod. Under Average Power Range

Monitor - Rod Block Monitor Technical

Specifications/Maximum Extended Load

Line Limit Analysis (ARTS/MELLLA) on

each selection of a control rod, the average

of the assigned, unbypassed LPRMs is

adjusted to equal a 100% reference signal for

each of the two RBM channels. Each RBM

channel automatically limits the local

thermal margin changes by limiting the

allowable change in local average neutron

flux to the RBM setpoint. If the local average

neutron flux change is greater than that

allowed by the RBM setpoint, within either

RBM channel, the rod withdrawal permissive

is removed preventing further rod movement.

Since the change in local neutron flux is

calculated from the change in the average of

the LPRM readings, and calibrated on every

rod selection to the reference signal, offsets

in individual LPRM readings due to

calibration differences are effectively

eliminated for a given RBM setpoint.

Therefore, the constraints on the withdrawal

of any given rod are unchanged, and there

will not be any increase in RWE delta-CPR.

Since the MCPR Safety Limit is unaffected

and the delta-CPR values are unchanged, the

cycle CPR Operating limits are unchanged

due to this TS change. Therefore, the

proposed change in the frequency of LPRM

signal calibration does not result in a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

Sr., Esquire, Sr. V. P. and General

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: May 3,

1996

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS

Surveillance Requirements 4.6.5.3.a and

4.6.5.4.a to modify specific

requirements to perform surveillance

flow testing of the Standby Gas

Treatment and Reactor Enclosure

Recirculation Systems 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. The proposed Technical Specifications

(TS) changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed TS changes do not involve

any physical changes to plant systems or

equipment. The proposed TS changes only

change the Surveillance Requirements (SRs)

surveillance test frequency pertaining to flow

testing of the SGTS and RERS from monthly

to quarterly. The periodic surveillance test

frequencies provide adequate assurance that

the equipment tested will remain in an

operable condition. The test frequency

interval for the flow testing of the SGTS and

RERS was determined from the regulatory

position in USNRC Regulatory Guide 1.52,

‘‘Design, Testing, and Maintenance Criteria

for Post Accident Engineered-Safety-Feature

Atmosphere Clean-up System Air Filtration

and Adsorption Units of Light-Water-Cooled

Nuclear Power Plants’’. As stated in

Regulatory Position C.4.d, ’’... each

Engineered Safety Feature (ESF) atmosphere

cleanup train should be operated at least 10

hours per month, with the heaters on (if so

equipped), in order to reduce the buildup of

moisture on the absorbers and HEPA filters.’’

System operation on a monthly basis for

the purpose of preventing moisture buildup

on the absorbers as described in R.G. 1.52 is

not required at Limerick due to the

continuous dry instrument air purge

described previously in the Safety

Assessment section of this submittal.

Therefore a change in the interval between

tests from monthly to quarterly will not

result in moisture accumulation which

would reduce the capability of the absorber

to remove the iodine species from the

exhaust air flow stream.

The SGTS components are common to both

units and must be run with the associated

RERS for the surveillance test for each unit.

The currently specified test frequency results

in the SGTS being run at least twice per

month or as many as eight (8) times per

quarter for this surveillance, in addition to

other required system surveillance tests

which require the use of the components in

this system. A change in surveillance test

frequency from monthly to quarterly would

reduce the wear on system components and

thereby reduce the associated system

downtime for maintenance and repairs. The

consequent increased availability provides

greater assurance that the system will be able

to perform its mitigation function following

any postulated accident.

Surveillance test frequency on a quarterly

interval is considered adequate to verify

operability, as demonstrated by the required

quarterly test interval for other equipment

important to safety which have a similar

function, such as the requirement for

quarterly verification of the isolation time of

the secondary containment and refueling area

isolation valves, as required by LGS TS

Sections 4.6.5.2.1 and 4.6.5.2.2.

Therefore, the proposed TS changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS changes only involve

changes to the frequency in which the

specified surveillances tests are performed.

The proposed TS changes do not physically

change the design or intended function of the

systems, structures, or components

associated with the SGTS or RERS. There

will be no change to the existing redundancy

of systems and components. The proposed

change in surveillance test frequency will not

introduce the possibility of any failure

mechanisms of a different type than those

already evaluated in the SAR. The existing

components will not be used in any different

manner and no new components will be

added. Therefore with no physical changes

and no new or different manner of system

operation, no new failure mechanisms or

equipment failure modes are created.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the LGS

TS Bases has not been reduced. The specific

basis for the 31 day surveillance interval is

not given in the LGS TS Bases section nor in

the LGS UFSAR Sections 6.5.1 or 9.4.2 which

discuss the subject systems. However,

Regulatory Position C.4.d of Regulatory

Guide 1.52, Revision 2, relating to

maintenance requirements, recommends:

≥Each ESF atmosphere cleanup train

should be operated at least 10 hours per

month, with the heaters on (if so equipped),

in order to reduce the buildup of moisture on

the absorbers and HEPA filters.’’
64392 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

The Bases for Surveillance Requirements

(SR) 3.6.4.3.1 in the Standard Technical

Specifications for General Electric Plants,

BWR/4, which corresponds to the subject

LGS TS test, also notes the need for ten (10)

hours of operation per month for elimination

of moisture in the filters.

The basis for the requirement for a monthly

test with the heaters energized is clearly

related to the desired elimination of moisture

in the filters and absorbers. However, LGS

UFSAR Table 6.5-2 states that LGS does not

conform to R.G. 1.52, Position C.4.d because

the SGTS and RERS trains are ‘‘continuously

purged with dry instrumentation air to

prevent build-up of moisture.’’ UFSAR

Sections 6.5.1.1.2 and 6.5.1.3.2 provide

additional discussion of this method of

moisture control.

Therefore, the proposed TS 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 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

Sr., Esquire, Sr. V. P. and General

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request:

September 27, 1996

Description of amendment request:

The proposed Technical Specifications

(TS) changes would increase the Reactor

Enclosure Secondary Containment

maximum inleakage rate. This change

will also impact secondary containment

drawdown time and system flow rate

assumptions, thereby, affecting charcoal

filter bed efficiency and post accident

dose analysis.

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 TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Changing the Reactor Enclosure post

drawdown inleakage rate from 1250 cfm to

2500 cfm does not involve any changes to the

function or operation of any plant component

or safety related system. The Reactor

Enclosure Recirculation System (RERS) and

the Standby Gas Treatment System (SGTS)

will maintain their design function by

mitigating the radiological consequences of

the analyzed accident and mitigating the post

LOCA temperatures within the Reactor

Enclosures. No analyzed accident initiating

events are impacted, no new accident

initiators are created, and no new failure

modes are created. There are no changes to

the redundancy, separation, quality

assurance or fire protection requirements for

the associated components and systems.

The proposed changes to the LGS adsorber

bed residence time will no longer fully meet

the literal design guidance provided in

Regulatory Guide (RG) 1.52, ‘‘Design, Testing,

and Maintenance Criteria for Post Accident

Engineered-Safety-Feature Atmosphere

Cleanup System Air Filter and Adsorption

Units of Light-Water-Cooled Nuclear Power

Plants,’’ Revision 2, March 1978. This is

because LGS’s unique, yet more conservative,

adsorber bed design is not addressed by the

RG residence time design guidance.

However, the LGS SGTS charcoal adsorbers

still conform to the design function described

in RG 1.52, based on the following: The LGS

design with increased inleakage will

continue to conform to the three conditions

specified by RG 1.52, Position C.6.a, in order

to maintain an assigned decontamination

efficiency of 99%; there is a conservative

amount of charcoal adsorber material

provided by the LGS design, based on

calculations performed in accordance with

RG 1.3 ‘‘Assumptions Used For Evaluating

The Potential Radiological Consequences of a

Loss of Coolant Accident For Boiling Water

Reactors; and the LGS charcoal bed design is

more conservative than the RG 1.52 design

guidance, based on data (i.e., Iodine

Penetration vs. Air Velocity) published by

the charcoal manufacturer.

Therefore, the probability of occurrence

and the consequences of a malfunction of

equipment important to safety is not

increased. Also, the probability of occurrence

of an accident previously evaluated is not

increased. However, the proposed changes do

affect the leak tightness of the Unit 1 and

Unit 2 Reactor Enclosure, which increases

the consequences of a postulated accident

previously evaluated.

Changing the Reactor Enclosure post

drawdown inleakage rate from 1250 cfm to

2500 cfm will result in an increase in the

calculated LOCA/LOOP Design Basis

Accident (DBA) off-site and on-site doses. 10

CFR Part 100, and 10 CFR Part 50 Appendix

A, General Design Criteria (GDC) 19,

establish reference dose values used to

determine site suitability and provide

reasonable assurance that the facility can be

operated following the analyzed accident

without undue risk to the health and safety

of the public. The proposed TS changes will

increase the SGTS drawdown time from 2

minutes and 20 seconds to 15 minutes and

30 seconds. The drawdown time increase

will not prevent the RERS/SGTS from

performing all of their safety related

functions. However, because it is

conservatively assumed that all radioactive

material released during the drawdown

period is unfiltered, and because the

drawdown period has been extended

whereby more unfiltered radioactive material

is assumed to be released following the DBA,

there is a corresponding increase in the

calculated Exclusion Area Boundary (EAB),

Low Population Zone (LPZ), and Control

Room doses. It is also assumed that the SGTS

exhausts at the maximum inleakage rate

throughout the entire DBA evaluation period

(i.e., 30 days) where an increase in the

maximum inleakage rate would also

contribute to higher postulated EAB, LPZ,

and Control Room doses. However, the

proposed calculated doses do not exceed 10

CFR Part 100, or 10 CFR Part 50, Appendix

A, DGC 19 reference doses.

Since the proposed doses resulting from

the changes remain below 10 CFR Part 100,

and 10 CFR Part 50, Appendix A, these

proposed changes will not significantly

increase the consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Changing the Reactor Enclosure post

drawdown inleakage rate from 1250 cfm to

2500 cfm is not an accident initiator nor does

it result in the occurrence of an accident. The

changes do not affect the function or

operation of any plant component or safety

related system nor do they create any new

failure modes.

In addition, the proposed changes do not

involve any changes to the function or

operation of any plant system or component

nor will they adversely affect the Reactor

Enclosure post LOCA environmental

conditions. Furthermore, these changes will

not create any new or different failure modes

for the equipment important to safety within

the Reactor Enclosure Secondary

Containment.

Therefore, the possibility of an accident of

a different type or a different type of

malfunction of equipment important to safety

than previously evaluated is not created.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

Changing the Reactor Enclosure post

drawdown inleakage rate from 1250 cfm to

2500 cfm will result in reducing the margin

of safety as defined in the LGS Updated Final

Safety Analysis Report (UFSAR) relative to

the off-site and on-site doses following a

LOCA/LOOP DBA, and an increase of the

UFSAR specified system drawdown time.

From a system perspective, increasing the

SGTS drawdown time from 2 minutes and 20

seconds to 15 minutes and 30 seconds will

not prevent the RERS/SGTS from performing

all of their safety related functions. There

will be a postulated increase in the

corresponding EAB, LPZ, and Control Room

doses, since it is assumed that fuel damage

occurs coincident with the LGS DBA (i.e, at

time = 0), all radioactive material released

during the drawdown time is unfiltered, and

the drawdown time is proposed to be

extended whereby more unfiltered

radioactive material could be released. It is

also assumed that the SGTS exhausts at the

maximum inleakage rate throughout the

entire DBA evaluation period (i.e., 30 days)

where an increase in the maximum inleakage
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64393

rate would also contribute to higher

postulated EAB, LPZ, and Control Room

doses. However, these calculated doses will

remain below 10 CFR Part 100, and 10 CFR

Part 50, Appendix A, GDC 19 reference

doses.

Therefore, these 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 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

Sr., Esquire, Sr. V. P. and General

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: October

1, 1996

Description of amendment request:

The proposed amendment would allow

for a one-time extension of the

surveillance intervals for the

containment isolation valve (CIV) seat

leakage test, the isolation valve seal

water test, the boron injection tank

leakage test, the containment spray

nozzle test, and the city water backup to

the auxiliary boiler feed pump test.

These tests would be performed during

the refueling outage scheduled to begin

in April 1997.

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:

Regarding the Containment Isolation Valve

seat leakage and Isolation Valve Seal Water

tests:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response:

The proposed amendment does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. The probability of a previously

evaluated accident will not increase because

CIV leakage does not provide any role in

accident initiation. The CIVs provide

containment isolation following a design

basis accident.

The consequences of an accident

previously evaluated will not significantly

increase because the CIV leakage

measurements contain significant margin to a

more restrictive criteria based on the

requested surveillance interval extension. As

discussed in Section II, ‘‘Evaluation of

Changes,’’ [see application dated October 1,

1996] based on an evaluation of past CIV leak

tests, the proposed change will not result in

an increase in containment leakage because

the measured leakage in previous CIV leak

tests shows large margin to a more restrictive

criteria based on the requested surveillance

interval extension. Also, the latest test of

IVSWS [isolation valve seal water system]

satisfied the established acceptance criteria.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response:

The proposed license amendment does not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed change only

provides for a relatively short, one-time

extension of the current leak-test interval for

certain containment isolation valves. The

proposed change does not involve the

addition of any new or different type of

equipment, nor does it involve operating

equipment required for safe operation of the

facility in a manner different from that

addressed in the Final Safety Analysis

Report. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

The proposed amendment does not involve

a significant reduction in a margin of safety.

The proposed change, for a one-time

extension of the test interval, will not result

in a significant reduction in a margin of

safety because the test interval is being

extended by only a short period and the

measured leakage in previous CIV leak tests

shows large margin to a more restrictive

criteria based on the surveillance interval

extension. In addition, the online leakage

monitoring capability of the WCCPPS [weld

channel containment penetration

pressurization system] helps ensure that

changes in CIV leakage during the extension

period will be detected. Therefore, this

change does not create a significant reduction

in a margin of safety.

Regarding the Boron Injection Tank (BIT)

leakage test:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed change

will not degrade the integrity of the BIT

piping outside containment because no time

dependent failure trends were observed in

the review of past test results. The

probability of a previously evaluated

accident will not be increased because BIT

leakage does not provide any role in accident

prevention. The BIT leakage test only verifies

that the BIT and associated piping meet

specified leakage limits.

The consequences of an accident

previously evaluated will not significantly

increase because the BIT leakage test results

show large margins to the allowable leakage

limit.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response:

The proposed license amendment does not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed change does no[t]

involve the addition of any new or different

type of equipment, nor does it involve

operating equipment required for safe

operation of the facility in a manner that’s

different from that addressed in the Final

Safety Analysis Report. Also, the increased

surveillance interval (one-time only) will not

adversely affect the integrity of the BIT

piping and will not result in any new failure

modes. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

The proposed license amendment does not

involve a significant reduction in a margin of

safety. Because of the large margin between

the previous test and the allowable leak rate

limit, the proposed change, for a one-time

extension of the test interval, for the BIT

leakage test does not adversely affect the

performance of any safety related system,

component, and does not result in increased

severity of any of the accidents considered in

the Final Safety Analysis Report. Based on

past test results, the one-time extension of

the leak test interval does not involve a

significant reduction in a margin of safety.

Regarding the Containment Spray Nozzle

test:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. As discussed in

Section II, ‘‘Evaluation of Changes,’’ [see

application dated October 1, 1996] based on

an evaluation of past test results the

proposed change will not degrade the

reliability of the Containment Spray Nozzles

because no time dependent failure trends

were observed in the data review. The

probability of a previously evaluated

accident will not be increased because the

Containment Spray Nozzles do not provide

any role in accident prevention. The

Containment Spray Nozzles provide a

uniform spray distribution for containment

cooling following postulated post-accident

conditions.

The consequences of an accident

previously evaluated will not increase

because the Containment Spray Nozzle

reliability is not degraded by this change.
64394 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response:

The proposed license amendment does not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed change does not

involve the addition of any new or different

type of equipment, nor does it involve

operating equipment required for safe

operation of the facility in a manner that is

different from that addressed in the Final

Safety Analysis Report. Also, the increased

surveillance interval (one-time only) w[i]ll

not adversely affect the functioning of the

Containment Spray Nozzles and will not

result in any new failure modes. Therefore,

the proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

The proposed license amendment does not

involve a significant reduction in a margin of

safety. The proposed change, for a one-time

extension of the test interval, for the

Containment Spray Nozzles does not

adversely affect the performance of any safety

related system, component, or instrument, or

safety system setpoints and does not result in

increased severity of any of the accidents

considered in the Final Safety Analysis

Report. Based on past test results, the onetime extension of the functional test interval

will not adversely affect the functioning of

the Containment Spray Nozzles. Therefore,

this change does not create a significant

reduction in a margin of safety.

Regarding the City Water Backup test:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed change

will not degrade the reliability of the City

Water Backup Supply Valves for the AFW

[auxiliary feedwater] System because no time

dependent failure trends were observed in

the review of past test results. The

probability of a previously evaluated

accident will not increase because the City

Water Backup Supply Valves for the AFW

System do not provide any role in accident

prevention. The City Water Backup Supply

Valves for the AFW System only provides a

diverse source of water for the AFW system.

The consequences of an accident

previously evaluated will not significantly

increase because the City Water Backup

Supply Valves for the AFW System are not

assumed to function to mitigate any analyzed

accident.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response:

The proposed license amendment does not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed change does not

involve the addition of any new or different

type of equipment, nor does it involve

operating equipment required for safe

operation of the facility in a manner that is

different from that addressed in the Final

Safety Analysis Report. Also, the increased

surveillance interval (one-time only) will not

adversely affect the functioning of the City

Water Backup Supply Valves for the ABFP

[auxiliary boiler feedpump] and will not

result in any new failure modes. Therefore,

the proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

The proposed amendment does not involve

a significant reduction in a margin of safety.

The proposed change, for a one-time

extension of the test interval, for the City

Water Backup Supply Valves for the ABFP

does not adversely affect the performance of

any safety related system, component, or

instrument, or safety system setpoints and

does not result in increased severity of any

of the accidents considered in the Final

Safety Analysis Report. Based on past test

results, the one-time extension of the

functional test interval will not adversely

affect the functioning of the City Water

Backup Supply Valves for the AFW System.

Therefore, this change does not create 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: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa,

Acting

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: October

25, 1996

Description of amendment request:

The proposed change to Hope Creek

Technical Specification (TS) 3/4.1.3.5,

‘‘Control Rod Scram Accumulator’’,

would: 1) permit a separate entry into a

Technical Specification action

statement for each inoperable control

rod; 2) provide more specific

applicability for required actions in

operational condition 1 or 2 with one

inoperable control rod scram

accumulator (reactor pressure of greater

than or equal to 900 psig would be

specified); 3) provide more specific

actions (verify charging water pressure)

for two or more inoperable control rod

scram accumulators and reactor

pressure is greater than or equal to 900

psig; 4) provide more specific actions

when reactor pressure is less than 900

psig and one or more control rod scram

accumulators are inoperable (verify

insertion of control rods associated with

inoperable accumulators and verify that

charging water header pressure is

greater than or equal to 940 psig); and

5) provide specific actions in

operational condition 5 with one or

more withdrawn control rods

inoperable; and 6) eliminate the

requirements to perform a 18-month

channel functional test of the leak

detectors and the 18-month channel

calibration of the pressure detectors.

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 change incorporates the appropriate

content of the improved BWR/4 Standard

Technical Specifications, NUREG-1433, for

Control Rod Scram Accumulators.

The proposed Technical Specification and

required Action completion times are

consistent with or more conservative than

those approved for use in the improved

Technical Specifications for inoperable

control rod scram accumulators. In addition,

the proposed surveillance requirements for

the control rod scram accumulators are

sufficient to adequately demonstrate

operability as stated in the Bases for the

improved Technical Specifications. Further,

the proposed changes enhance the current

Hope Creek Technical Specifications by

reflecting improved techniques collectively

learned by the industry. Therefore, the

proposed changes do not significantly

increase the risk or consequences of any

accidents previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Neither the mechanism for initiating or

completing a scram is modified by this

proposed change. There are no physical

changes to plant equipment proposed in the

application. The proposed change does not

create a means by which the scram function

could be impeded or prevented. The

proposed change is functionally equivalent to

the current Technical Specifications, but

provides additional operational flexibility to

diagnose and resolve equipment issues that

do not impact operability of the control rods

before taking proscriptive actions which
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64395

result in significant plant transients (i.e. full

power scram).

3. The proposed change does not involve

a significant reduction in a margin of safety.

The operability of the accumulators and

the scram function of the control rod drive

system protects the Safety Limit Minimum

Critical Power Ratio as well as the 1%

cladding plastic strain fuel design limit. The

proposed change does not reduce a margin of

safety as defined in the Bases of the

Technical Specification since the proposed

change does not affect the maximum

allowable scram times for control rods, nor

does it change the maximum allowable

number or minimum separation of inoperable

control rods. The proposed change does not

modify any instrument setpoints or

functions. The proposed change will either

maintain the present margins of safety or

increase them, by reducing the need for

unnecessary challenges to the reactor

protection system and resulting plant

shutdowns, while still maintaining the

capability to complete a reactor scram.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Rochester Gas and Electric

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

Ginna Nuclear Power Plant, Wayne

County, New York

Date of amendment request: October

29, 1996

Description of amendment request:

The proposed amendment would revise

the mode of applicability for the motordriven auxiliary feedwater (AFW) pump

actuation on opening of the main

feedwater (MFW) pump breakers to

correct an error introduced during

Amendment No. 61.

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 less restrictive changes discussed in

Section C.1 [of the licensee’s application] do

not involve a significant hazards

consideration as discussed 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 correct an error which was introduced

in Amendment No. 61 to the Ginna Station

technical specifications. The changes revert

the mode of applicability for the motordriven AFW pump actuation on the opening

of the MFW pump breakers to what existed

previously. The change is essentially

correction of a typographical error that was

caused through use of the electronic version

of NUREG-1431 in preparation of the Ginna

Station ITS [Improved Technical

Specifications]. There have been no

subsequent plant modifications or changes to

the accident analysis which would invalidate

the previous NRC acceptance of only

requiring this Function above 5% power. The

accident analyses do not credit automatic

initiation of AFW on MFW pump trip in

MODE 2. As such, these changes 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 which

existed prior to Amendment No. 61. 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

there have been no subsequent plant

modifications or changes to the accident

analysis which would invalidate the previous

NRC acceptance of only requiring this

Function above 5% power. 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: S. Singh Bajwa,

Acting

Rochester Gas and Electric

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

Ginna Nuclear Power Plant, Wayne

County, New York

Date of amendment request: October

29, 1996

Description of amendment request:

The proposed amendment would revise

the Required Actions for the auxiliary

feedwater (AFW) pump actuation on

Steam Generator Level (SG) - Low Low

logic to be consistent with those

specified in NUREG-1431.

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 less restrictive changes discussed in

Section C.1 [of the licensee’s application] do

not involve a significant hazards

consideration as discussed 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

with respect to the Required Actions for

AFW actuation on SG Level - Low Low logic

provide consistency with NUREG-1431 by

requiring an inoperable channel to be placed

in the tripped condition within 6 hours. The

affected logic then requires 1 of 2 channels

in order to actuate such that there is no

impact on any 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 AFW actuation on SG Level - Low Low

still remains capable of performing its

function with an inoperable channel placed

in the tripped configuration. These changes

are also consistent with those provided in

NUREG-1431. 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
64396 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

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: S. Singh Bajwa,

Acting

Virginia Electric and Power Company,

Docket Nos. 50-338 and 50-339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

September 4, 1996

Description of amendment request:

The proposed amendment to the

Technical Specifications would allow

the use of four lead test assemblies

(advanced zirconium-based alloys) in

the North Anna, Units 1 and 2, reactor

cores.

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 the four FCF [Framatome

Cogema Fuels] lead test assemblies will not:

1.Involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The FCF lead test assemblies are

very similar in design to the Westinghouse

fuel that comprises the remainder of the core.

The reload core design for North Anna cycles

which incorporate the lead test assemblies

will meet all applicable design criteria. In

addition, the performance of the ECCS

[emergency core cooling system] at North

Anna Units 1 and 2 will not be affected by

the insertion of the four lead test assemblies,

so the criteria of 10 CFR 50.46 will be

satisfied for use of these assemblies with fuel

rods, guide thimble tubes, and

instrumentation tubes fabricated with

advanced zirconium-based alloys. The use of

these fuel assemblies will not result in a

change to the North Anna Units 1 and 2

reload design and safety analysis limits. The

existing safety analyses based on the resident

Westinghouse fuel will remain applicable for

cycles which incorporate the lead test

assemblies. Therefore, neither the probability

of occurrence nor the consequences of any

accident previously evaluated is significantly

increased.

2. Create the possibility for a new or

different type of accident from any accident

previously evaluated. The FCF lead test

assemblies are very similar in design (both

mechanical and composition of materials) to

the resident Westinghouse fuel. North Anna

cores which incorporate the lead test

assemblies will be designed to meet all

applicable design criteria and ensure that all

pertinent licensing basis criteria are met.

Demonstrated adherence to these standards

and criteria precludes new challenges to

components and systems that could

introduce a new type of accident. North

Anna safety analyses based on the resident

Westinghouse fuel will remain applicable for

cores containing the lead test assemblies. All

design and performance criteria will

continue to be met and no single failure

mechanisms have been created. In addition,

the use of these fuel assemblies does not

involve any alteration to plant equipment or

procedures which would introduce any new

or unique operational modes or accident

precursors. Therefore, the possibility for a

new or different kind of accident from any

accident previously evaluated is not created.

3. Involve a significant reduction in the

margin of safety. The use of the FCF lead test

assemblies does not change the performance

requirements on any system or component

such that any design criteria will be

exceeded, and will not cause the core to

operate in excess of pertinent design basis

operating limits. North Anna reload core

designs for cycles which incorporate the lead

test assemblies will specifically evaluate any

pertinent differences between the lead test

assemblies and the resident fuel, and will

take into consideration the normal core

operating conditions allowed in the

Technical Specifications. Safety analyses

based on the resident Westinghouse fuel will

remain applicable for cores incorporating the

FCF lead test assemblies. Analyses or

evaluations will be performed each cycle to

confirm that the criteria in 10 CFR 50.46 will

be met. Therefore, the margin of safety as

defined in the Bases to the North Anna Units

1 and 2 Technical Specifications is not

significantly reduced.

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: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Acting Project Director: Mark

Reinhart

Virginia Electric and Power Company,

Docket Nos. 50-338 and 50-339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

November 6, 1996

Description of amendment request:

The proposed changes will modify the

requirements for isolated loop startup to

permit filling of a drained isolated loop

via backfill from the reactor coolant

system through partially open stop loop

valves.

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:

Specifically, operation of the North Anna

Power Station [in] accordance with the

proposed changes will not:

1. Involve a significant increase in the

probability of occurrence or consequences of

an accident previously evaluated. The

probability of occurrence of a positive

reactivity addition accident is not being

increased by the proposed Technical

Specification change. The proposed

restrictions on boron concentration and

mixing, reactor coolant system inventory and

reactivity and count rate monitoring provide

a level of protection against reactivity

addition accidents which is equivalent to

that currently in place.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated. The proposed change

does not introduce any new or unique failure

modes or accident precursors. Eliminating

the operability requirements for the loop stop

valve interlocks does not create any new or

different kind of accident scenario. Loop

startup accidents in the various modes of

operation have been analyzed. Operation of

the loop stop valves will not change. New

requirements have been imposed for the case

of backfilling a drained loop from the reactor

coolant system to ensure that core cooling

and reactivity control are preserved

throughout the backfill evolution.

3. Involve a significant reduction in any

margin of safety. The new Technical

Specification loop isolation and startup

requirements for temperature, boron

concentration, and shutdown margin fulfill

the function of the loop stop valve interlocks.

Therefore, the margin of safety as defined in

any Technical Specification bases is not

reduced.

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: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: Mark Reinhart

(Acting)
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64397

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: October

31, 1996

Description of amendment request:

The proposed amendment would revise

the Kewaunee Nuclear Power Plant

(KNPP) Technical Specifications (TS) by

deleting the requirement for an annual

submittal of a description of changes

made pursuant to 10 CFR 50.59.

Consistent with 10 CFR 50.59(b)(2), a

description of changes will

subsequently be included with the

KNPP Updated Safety Analysis Report

(USAR) update in accordance with 10

CFR 50.71(e). Additionally, the

proposed amendment would correct

minor administrative inconsistencies in

the TS Table of Contents and in a

footnote reference.

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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c)The

NRC staff’s review is presented below:

On August 31, 1992 (57 FR 39353),

the NRC amended 10 CFR 50.59(b)(2) to

reduce the regulatory burden on nuclear

licensees. This action revised the

requirements for the annual submission

of reports for facility changes under 10

CFR 50.59. This action did not affect the

substance of the evaluation or the

documentation required for 10 CFR

50.59 type changes. It only affected the

interval for submission of the

information to the NRC. Instead of

submitting the information annually,

the information can be submitted on a

refueling cycle basis, provided the

interval between successive reports does

not exceed 24 months.

In order to take advantage of this

reduction in regulatory burden, the

licensee has proposed an amendment to

remove the submittal of a report of

facility changes under 10 CFR 50.59

from the Technical Specification list of

annual reporting requirements.

Additionally, the licensee has proposed

corrections to minor administrative

inconsistencies in the TS Table of

Contents and in a footnote reference.

The proposed changes are

administrative only and do 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.

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 Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P. O.

Box 1497, Madison, Wisconsin 53701-

1497

NRC Project Director: Gail H. Marcus

NOTICE OF ISSUANCE OF

AMENDMENTS TO FACILITY

OPERATING LICENSES

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power

Station,Plymouth County,

Massachusetts

Date of application for amendment:

May 1, 1996, as supplemented August

12, 1996.

Brief description of amendment: The

amendment approves relocation of the

administrative controls related to the

quality assurance review and audit

requirements of Section 6, Technical

Specifications 6.5.B.8, ‘‘Nuclear Safety

Review and Audit Committee-Audits,’’

from the Pilgrim Station Technical

Specifications to the Boston Edison

Quality Assurance Manual (BEQAM).

This change is in accordance with the

guidance contained in NRC

Administrative Letter 95-06,

‘‘Relocation of Technical Specification

Administrative Controls Related to

Quality Assurance.’’ In addition, the

Safety Evaluation includes the NRC staff

review and approval of the BEQAM

changes in support of this amendment.

Date of issuance: November 12, 1996

Effective date: November 12, 1996

Amendment No.: 168

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28605)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 12, 1996.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Commonwealth Edison Company,

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

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

August 29, 1996, as supplemented on

September 20, 1996, and October 4,

1996.

Brief description of amendments: The

amendments 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 an

exception as detailed in the licensee’s

application.

Date of issuance: November 12, 1996

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 175 and 162
64398 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

Facility Operating License Nos. DPR39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 9, 1996 (61 FR 52964).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated November 12, 1996.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

August 14, 1996, as supplemented

October 18, 1996, and related

application of January 18, 1996

Brief description of amendment: The

amendment revises the technical

specifications (TS) to allow one-cycle

deferral of the inspection of reactor

coolant pump (RCP) flywheels.

Date of issuance: November 7, 1996

Effective date: November 7, 1996

Amendment No.: 175

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 24, 1996 (61 FR

50054) The October 18, 1996, letter

provided an updated TS page. This

change was within the scope of the

original application and did not change

the staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 7, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

December 14, 1994, as supplemented by

letters dated May 16 and August 29,

1996

Brief description of amendments: The

amendments will incorporate guidance

and recommendations for diesel

generators contained in NUREG-1366,

‘‘Improvements to Technical

Specifications Surveillance

Requirements,’’ Generic Letter (GL) 93-

05, ‘‘Line-Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operations,’’ GL 94-01, ‘‘Removal of

Accelerated Testing and Reporting

Requirements for Emergency Diesel

Generators,’’ and NUREG-1431,

‘‘Revised Standard Technical

Specifications for Westinghouse PWRs.’’

Date of issuance: November 12, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 170 and 152

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

Technical Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28612)

The August 29, 1996, letter provided

clarifying information that did not

change the scope of the December 14,

1996, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 12,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50-458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: August 1,

1996

Brief description of amendment: The

amendment revises the technical

specifications to incorporate

requirements for limiting the time that

the hydrogen mixing isolation valves on

the drywell are open. The amendment

also changes the time from 7 days to 31

days to determine the cumulative time

the valves are open.

Date of issuance: November 12, 1996

Effective date: November 12, 1996

Amendment No.: 89

Facility Operating License No. NPF47. The amendment revised the

Technical Specifications/operating

license.

Date of initial notice in Federal

Register: September 25, 1996 (61 FR

50343) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 12, 1996.No significant

hazards consideration comments

received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Entergy Operations, Inc., System

Energy Resources, Inc.,

SouthMississippi Electric Power

Association, and Entergy Mississippi,

Inc., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

May 9, 1996, as supplemented by letter

dated August 27, 1996.

Brief description of amendment: The

amendment changed Surveillance

Requirements (SRs) 3.4.4.3, Safety/

Relief Valves, 3.5.1.7, Automatic

Depressurization System Valves, and

3.6.1.6.1, Low-Low Set Valves, of the

Technical Specifications and allows the

licensee to perform the surveillance of

the relief mode of operation of the

safety/relief valves on the main steam

lines without physically lifting the disk

of a valve off the seat at power. The

changes stated that the required

operation of the valve to verify is that

the relief-mode actuator strokes when

the valve is manually actuated and the

frequency of the surveillances are in

accordance with the inservice testing

program for the valves.

Date of issuance: November 18, 1996

Effective date: November 18, 1996

Amendment No: 130

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47971) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 18, 1996. No significant

hazards consideration comments

received: No

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. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

June 3, 1996, as supplemented October

23, 1996

Brief description of amendment: The

amendment clarifies a restriction on

shutdown margin monitor operability

while changing operational modes, so

that it only limits reactivity changes

caused by boron dilution and rod

withdrawal. The amendment also

corrects a technical specification

numerical reference so that the

specification number cited is in

agreement with Amendment 99, dated

December 29, 1994.

Date of issuance: November 14, 1996
Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices 64399

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 131

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 20, 1996 (61 FR 31559)

The October 23, 1996, letter provided

clarifying information that did not

change the scope of the June 3, 1996,

application and the initial proposed no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

November 14, 1996No 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.

Power Authority of the State of New

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

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

May 30, 1996

Brief description of amendment: The

proposed change to the anticipated

transient without scram recirculation

pump trip logic for the James A.

Fitzpatrick Nuclear Power Plant allows

for a high pressure trip setpoint which

is dependent upon the number of safety/

relief valves which are out of service.

Date of issuance: November 7, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 237

Facility Operating License No. DPR59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34896) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated November 7, 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.

Power Authority of the State of New

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

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

May 30, 1996, as supplemented October

17, and November 8, 1996

Brief description of amendment: The

proposed amendment changes the

FitzPatrick safety limit minimum

critical power ratio from its current

value of 1.07 for two recirculation loop

operation to 1.09 and from 1.08 to 1.10

for single recirculation loop operation

for the Cycle 13 operation.

Date of issuance: November 14, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 238

Facility Operating License No. DPR59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34896) The

October 17 and November 8, 1996

letters provided supplemental

information that did not change the

initial no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated November 14, 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.

Southern Nuclear Operating Company,

Inc., Docket No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: August

23, 1996, as supplemented by letters

dated September 16, November 6, 11

and 14, 1996

Brief description of amendment: The

amendment changes the Technical

Specifications (TS) to allow installation

of laser welded elevated tubesheet

sleeves. Specifically, the amendment is

for one cycle only for Farley Unit 2.

Permanent, generic TS changes for

Westinghouse laser welded sleeves for

both units will be submitted prior to the

next Unit 1 refueling outage currently

scheduled for spring 1997.

Date of issuance: November 20, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment No.: 117

Facility Operating License No. NPF-8:

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47982) The September 16, November 6,

11 and 14, 1996, letters provided

clarifying information that did not

change the scope of the August 23,

1996, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 20,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

July 17, 1995.

Brief description of amendments:

These amendments revise the frequency

of surveillance requirements for certain

plant protective system instrumentation

contained in Technical Specifications

(TS) 3.3.1, ‘‘Reactor Protective System

(RPS) Instrumentation - Operating,’’ TS

3.3.2, ‘‘Reactor Protective System (RPS)

Instrumentation - Shutdown,’’ TS 3.3.3,

‘‘Control Element Assembly Calculators

(CEACs),’’ TS 3.3.4, ‘‘Reactor Protective

System (RPS) Logic and Trip Initiation,’’

TS 3.3.5, ‘‘Engineered Safety Features

Actuation System (ESFAS)

Instrumentation,’’ and TS 3.3.6,

‘‘Engineered Safety Features Actuation

System (ESFAS) Logic and Manual

Trip.’’

Date of issuance: November 18, 1996

Effective date: November 18, 1996, to

be implemented within 30 days of the

date of issuance.

Amendment Nos.: Unit 2 - 133 ; Unit

3 - 122

Facility Operating License Nos. NPF10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45185)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated November 18, 1996.No

significant hazards consideration

comments received: No.Temporary

Local Public Document Room

location: Science Library, University of

California, P. O. Box 19557, Irvine,

California 92713

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 application for amendment:

September 4, 1996

Brief description of amendment: This

amendment revises Technical
64400 Federal Register / Vol. 61, No. 234 / Wednesday, December 4, 1996 / Notices

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.

Date of issuance: November 8, 1996

Effective date: November 8, 1996, to

be implemented not later than 90 days

after issuance

Amendment No.: 212

Facility Operating License No. NPF-3:

Amendment revised the Technical

Specifications. Date of initial notice in Federal

Register: October 9, 1996 (61 FR 52970)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 8, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

July 18, 1996 Brief description of amendment: The

amendment adopts ASTM D-3803-1989

as the laboratory testing standard for

charcoal samples from the charcoal

adsorbers in the auxiliary/fuel building

emergency exhaust system. Date of issuance: November 13, 1996

Effective date: November 13, 1996, to

be implemented within 30 days of the

date of issuance.

Amendment No.: 118

Facility Operating License No. NPF30: The amendment revised the

Technical Specifications. Date of initial notice in Federal

Register: August 14, 1996 (61 FR 42285)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 13, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Dated at Rockville, Maryland, this 26th day

of November 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II

Office of Nuclear Reactor Regulation

[Doc. 96–30714 Filed 12–3–96; 8:45 am]

BILLING CODE 7590–01–F

NUCLEAR WASTE TECHNICAL

REVIEW BOARD

Pahrump, Nevada: Yucca Mountain

Testing and Exploration Program,

Environmental Impact Statement,

Interim Storage Studies,

Transportation Infrastructure, Options

for Reducing Hydrogeologic

Uncertainties in the Proposed

Repository Waste Emplacement Area,

and Performance Assessment Issues;

Board Meeting

Pursuant to its authority under

section 5051 of Public Law 100–203, the

Nuclear Waste Policy Amendments Act

of 1987, the Nuclear Waste Technical

Review Board will hold its winter

meeting on Tuesday and Wednesday,

January 28–29, 1997, in Pahrump,

Nevada. The meeting will be held at the

Bob Ruud Community Center, 150 N.

Highway 160, Pahrump, Nevada 89048;

Tel (702) 727–9991. Sleeping

accommodations are available in the

Saddle West Hotel, 1220 S. Highway

160, Pahrump, Nevada 89048; Tel (702)

727–1111; Fax (702) 727–5315. To

receive the preferred rate, reservations

must be made by December 27, 1996.

The meeting is open to the public and

will begin at 8:30 a.m. both days. On the first morning, the Board will

hear presentations by representatives of

the Department of Energy (DOE) and its

contractors on the exploration and

testing program at Yucca Mountain,

Nevada; plans for preparing the

environmental impact statement for

repository development; and generic

studies on the development of an

interim spent fuel storage facility. The

Board is particularly interested in

hearing about how long it would take to

construct such a facility and to develop

a transportation infrastructure to move

significant quantities of waste. The

Board plans to invite Nye County

representatives to briefly summarize the

results of their scientific investigations

at the Yucca Mountain site.

The afternoon session will examine

the issues associated with DOE plans to

reduce, by late 1998, the current

uncertainties about the movement of

moisture through the proposed

repository waste emplacement area. The

focus will be on options for gathering

additional data, including what data

would be sought, and how the data

would be obtained.

On the second day of the meeting, the

morning session will address the

transportation of waste to a potential

repository at the Yucca Mountain site,

including an update on the DOE’s recent

privatization initiative and on more

local issues such as route selection and

emergency preparedness. The Board

plans to invite representatives from

Nevada state and local governments,

industry associations, and public

interest groups to make presentations. A

roundtable discussion will cover key

topics raised during the presentations.

The afternoon session will be devoted

to a discussion of performance

assessment. The Board has asked for

presentation on the DOE’s newly drafted

siting guidelines, 10 CFR 960, including

the basis for the proposed revisions. The

Board also would like to know about

DOE plans to make the logic and

reasoning that underlie performance

assessment ‘‘transparent’’ to both

scientific and lay communities.

Time has been set aside for public

comment and questions on both days.

To ensure that everyone wishing to

speak is provided time to do so, the

Board encourages those who have

comments to sign the Public Comment

Register, which will be located at the

registration table. A time limit may have

to be set on the length of individual

remarks; however, written comments of

any length may be submitted for the

record.

The Nuclear Waste Technical Review

Board was created by Congress in the

Nuclear Waste Policy Amendments Act

of 1987 to evaluate the technical and

scientific validity of activities

undertaken by the DOE in its program

to manage the disposal of the Nation’s

spent nuclear fuel and defense highlevel waste. In the same legislation,

Congress directed the DOE to

characterize a site at Yucca Mountain,

Nevada, for its suitability as a potential

location for a permanent repository for

the disposal of that waste.

Transcripts of this meeting will be

available via e-mail, on computer disk,

or on a library-loan basis in paper

format from Davonya Barnes, Board

staff, beginning February 26, 1997. For

further information, contact Frank

Randall, External Affairs, 1100 Wilson

Boulevard, Suite 910, Arlington,

Virginia 22209; (Tel) 703–235–4473;

(Fax) 703–235–4495.

Dated: November 22, 1996.

William Barnard,

Executive Director, Nuclear Waste Technical

Review Board.

[FR Doc. 96–30882 Filed 12–3–96; 8:45 am]

BILLING CODE 6820–AM–M