6402 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

initial scoping process, NRC issued a

scoping report in September 1998.

NRC’s initial scoping process was

based on the description of the PFSF

contained in the applicant’s submittal of

June 20, 1997, which did not include

the proposed rail line on public land

administered by BLM. This rail line

proposal was submitted to NRC on

August 28, 1998, as an amendment to

the PFS application. Similarly, BIA’s

conditional approval of the proposed

lease agreement was issued prior to the

applicant’s proposal of the rail line.

As a result of the applicant’s August

28, 1998, revision of its transportation

proposal, NRC, BIA, and BLM

determined that additional scoping

meetings should be conducted.

Additional scoping meetings were held

on April 29, 1999, in Salt Lake City, and

Tooele City, Utah. The meetings were

noticed in the Federal Register on April

14, 1999 (64 FR 18451). Primarily, the

scoping meetings focused on

environmental issues associated with

the rail line proposed in the applicant’s

August 28, 1998, license application

amendment, the request for issuance of

a ROW over public lands managed by

BLM, and environmental concerns

associated with the proposed lease

agreement that may not have been

addressed in the NRC’s initial scoping

process. In addition, interested parties

were also provided the opportunity to

submit written comments. Following

the additional scoping meetings and

comment period, a supplemental

scoping report was issued in November

1999.

Although STB was not identified as a

cooperating agency during the scoping

process, the environmental issues

related to its federal action (i.e.,

approving the construction and

operation of the proposed rail line) were

discussed during the scoping process.

STB has determined that these scoping

activities provided sufficient

opportunity for the public to comment

on the proposed action and the scope of

the EIS. Interested parties will have an

opportunity to provide comments on the

draft EIS.

Dated at Rockville, Maryland, this 27th day

of January 2000.

For the Nuclear Regulatory Commission.

E. William Brach,

Director, Spent Fuel Project Office, Office of

Nuclear Material Safety and Safeguards.

[FR Doc. 00–2930 Filed 2–8–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

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 January 14,

2000, through January 28, 2000. The last

biweekly notice was published on

January 26, 2000 (65 FR 4268).

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 March 10, 2000, 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 electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room). 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

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6403

Chairman of the Atomic Safety and

Licensing Board Panel, will rule on the

request and/or petition; and the

Secretary or the designated Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

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. 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 electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room).

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Dated of amendments request:

January 25, 2000.

Description of amendments request:

The proposed amendment requests a

revision to the definition of Response

Time Testing (RTT) for the Reactor

Protective System (RPS) and Engineered

Safety Features Actuation System

(ESFAS). The revision allows use of

either an allocated sensor response time

or a measured sensor response time for

pressure sensors used in channels of

RPS and ESFAS. The request is based

on Combustion Engineering NPSD–

1167, Revision 1, ‘‘Elimination of

Pressure Sensor Response Time Testing

Requirements—CEOG Task 1070.’’

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 licensing basis change

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated in the safety analysis

report.

This change to the licensing basis does not

result in a condition where the design,

material, and construction standards that

were applicable prior to the change are

altered. The same Reactor Protective System

and Engineered Safety Features Actuation

System instrumentation is being used; the

time response allocations/modeling

assumptions in Updated Final Safety

Analysis Report Chapter 14 analyses remain

the same; only the method of verifying time

response is changed. The proposed change

will not modify any system interface and

could not increase the likelihood of an

accident since these events are independent

of this change. The proposed activity will not

change, degrade or prevent actions or alter

any assumptions previously made in

evaluating the radiological consequences of

an accident described in the Updated Final

Safety Analysis Report. Therefore, the

proposed amendment does not result in any

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed licensing basis change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated in the safety analysis

report.

This change does not alter the performance

of the pressure and differential pressure

sensors used in the plant protection systems.

These sensors will still have their response

time verified before they are placed in

operational service and after any

maintenance to them that could affect their

response time. Changing the method of

periodically verifying instrument response

for certain sensor (assuring equipment

6404 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

operability) from time response testing to

calibration, use of actual data, and channel

checks will not create any new accident

initiators or scenarios. Periodic surveillance

of these instruments will detect significant

degradation in the sensor response

characteristic. Implementation of the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed licensing basis change

does not involve a significant reduction in

margin of safety.

The total Reactor Protective System and

Engineered Safety Features Actuation System

response time assumed in the safety analysis

is not affected by this change. The periodic

system response time verification method for

selected pressure and differential pressure

sensors is modified to allow the use of

allocated data based on actual test results or

other verifiable response time data.

Verification methods and calibration tests

assure that any degradation sufficient to

significantly affect sensor response time will

be detected before the total system response

time exceeds that defined in the safety

analysis. Therefore, it is concluded that the

proposed change does not result in a

significant reduction in margin with respect

to plant 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

amendments request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Marsha

Gamberoni, Acting.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Dated of amendment request:

November 18, 1999.

Description of amendment request:

The proposed amendment would

remove license condition 3.H, ‘‘Long

Term Program,’’ from Facility Operating

License DPR–35 for the Pilgrim Nuclear

Power Station.

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. No physical changes to

the facility will occur as a result of this

amendment. Work activities will continue to

receive the appropriate level of review in

accordance with Pilgrim procedures and

practices. The organizational structure and

processes that control and manage these

activities ensure activities are prioritized and

performed in a manner consistent with plant

safety. The proposed amendment removes an

administrative burden that is no longer

required.

(2) The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. No changes to the

physical design and operation of the plant

will occur as a result of this amendment. The

processes by which activities are planned,

prioritized, and controlled are not affected.

The appropriate level of technical review and

management oversight will continue to be

performed in accordance with existing

procedures and practices to ensure activities

are performed in a manner consistent with

plant safety.

(3) The proposed amendment does not

involve a significant reduction in a margin of

safety. As stated earlier, no changes to the

physical design and/or operation of any plant

systems will occur as a result of this

amendment; therefore, there is no reduction

in any margins of safety. Work activities will

continue to receive the appropriate technical

review and management oversight to ensure

activities are prioritized and performed in a

manner consistent with plant safety. The

proposed amendment removes an

administrative burden that is no longer

required.

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.

Attorney for licensee: W. S. Stowe,

Esquire, Entergy Nuclear Generation

Company, 800 Boylston Street, 36th

Floor, Boston, Massachusetts 02199.

NRC Section Chief: James W. Clifford.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Dated of amendment request:

November 29, 1999

Description of amendment request:

The proposed amendment would

relocate the requirements associated

with the high-steam-generator-level trip

functions of the Reactor Protective

System from the Technical

Specifications to the Technical

Requirements Manual.

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:

Criterion 1—Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated

The Steam Generator Level—High function

of the RPS [Reactor Protection System] is not

credited in any accident analyses nor does it

correspond to any TS [Technical

Specification] Safety Limit. The high-level

function acts to protect the Main Turbine

from excessive moisture carryover during

feedwater transient events. Protection of the

Main Turbine is not required to adequately

assure continued reactor safety or the health

and safety of the public. Although this

function may also serve to limit water

intrusion into the main steam lines and

consequential overcooling events, its role in

this capacity is insignificant, as it does not

directly act to secure feedwater from the

steam generators. This Steam Generator

Level—High function acts only to isolate the

Main Turbine from the steam generators by

causing a reactor trip, which in turn actuates

a turbine trip. This function does not meet

any of the criterions listed in 10 CFR

50.36(c)(2) (ii) for inclusion into the technical

specifications for ANO–2 [Arkansas Nuclear

One, Unit 2], and, therefore, may be excluded

from the TSs. Since no changes are made that

affect the current operation of this function

during its relocation to the ANO–2 TRM

[Technical Requirements Manual], and

because this function is not credited in any

accident analyses, no increase in the

probability or consequences of an accident

previously evaluated is evident.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

Criterion 2—Does Not Create the Possibility

of a New or Different Kind of Accident From

Any Previously Evaluated

The proposed changes relocate affected TS

requirements associated with the Steam

Generator Level—High Functions of the RPS

from the ANO–2 TSs to the ANO–2 TRM.

Future revisions to the setpoints and values

associated with this function will be

established within the requirements of 10

CFR 50.59 to ensure that excessive moisture

carryover is prevented in order to protect the

turbine and steam line loads. The Steam

Generator Level—High Trip setpoint is not

credited in any accident analyses and

performs only an equipment protection

function. The setpoint continues to protect

the Main Turbine from damage and preserves

operating margin to accommodate excessive

feedwater flow prior to trip.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3—Does Not Involve a Significant

Reduction in the Margin of Safety

The Steam Generator Level—High Trip

setpoint is not credited in any accident

analyses and performs only an equipment

protection function. The setpoint continues

to protect the Main Turbine from damage and

preserves operating margin to accommodate

excessive feedwater flow prior to trip. In

addition, turbine failure has been previously

evaluated at ANO–2 as not to be a significant

threat to the health and safety of the public.

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6405

Events that may result from water intrusion

into the main steam lines have been

previously evaluated and found not to rely

upon the Steam Generator Level—High Trip

function. The relocation of the requirements

associated with the Steam Generator Level—

High function from the TSs to the ANO–2

TRM does not change the current values and

requirements. Since no technical change in

the setpoint or allowable value is proposed

by this submittal and because the Steam

Generator Level—High function does not

meet any of the four criterion of 10 CFR

50.36(c)(2)(ii), no significant change to the

margin of safety is evident.

Therefore, this 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.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Dated of amendment request:

November 29, 1999 .

Description of amendment request:

The proposed amendment would revise

selected Technical Specifications (TSs),

Bases, and portions of the Safety

Analysis Report (SAR) to maintain

consistency with the transient and

accident analyses which evaluated the

impact of the replacement steam

generators (SGs) that are being used for

Cycle 15 operation. TS changes are

proposed for the Reactor Protection

System (RPS) and Engineered Safety

Features Actuation System (ESFAS) low

pressurizer pressure setpoints, the RPS

and ESFAS low SG pressure setpoints,

the RPS and ESFAS low SG level

setpoints, the reactor coolant flow rate

limit, and the high linear power trip

setpoints with inoperable main steam

safety valves (MSSVs). SAR changes

would support the new TS values and

would also include small increases in

calculated offsite radiological doses

using newer, more conservative

methods, for some non-loss-of-coolant

accident events. The doses would

remain within the 10 CFR Part 100

acceptance criteria. The proposed

amendment would also make changes to

the TSs and Bases that are not directly

related to the replacement SGs. These

changes would revise the allowed

outage time of the MSSVs in Modes 1

and 2 to allow up to 12 hours to reduce

the high linear power level-high trip

setpoint when one or more MSSVs are

inoperable, and would revise the action

statement in Mode 3 to maintain at least

two MSSVs operable on each SG.

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:

Criterion 1—Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated

The proposed changes to the ANO–2

[Arkansas Nuclear One, Unit 2] TSs are

analytically based which change setpoints

and procedure limits. No physical

modifications are required as a result of the

proposed changes. The RPS/ESFAS setpoint

changes provide functionally equivalent

protection with the RSGs [replacement steam

generators] as the previous setpoint values

provided with the OSGs [original steam

generators]. Proposed changes in regard to

RCS [reactor coolant system] flow rate and

High Linear Power Trip setpoints associated

with conditions where MSSVs are inoperable

represent appropriate restrictions that have

resulted from the various analyses performed

in support of RSG installation. An Emergency

Core Cooling System (ECCS) performance

analysis was performed to demonstrate

conformance to 10 CFR 50.46 for operation

with RSGs. For the large break Loss of

Coolant Accident (LOCA), the most limiting

single failure of the ECCS [is no failure to the

ECCS]. The small break LOCA analysis was

reanalyzed using the existing Supplement 2

Model (S2M) of the ABB CENP [ABB

Combustion Engineering Nuclear Power]

small break LOCA evaluation model. The

analysis was performed for 0.03 ft2, 0.04 ft2,

and 0.05 ft2 in the reactor coolant pump

(RCP) discharge leg. The results of both

analyses demonstrate continued conformance

to the ECCS acceptance criteria of 10 CFR

50.46. Non-LOCA analyses intended to

confirm the Chapter 15 events in the ANO–

2 SAR were also performed. The analyses

were performed considering the proposed

Safety Limits and the Limiting Safety

Settings of the TSs and were confirmed to be

bounding for the affected safety analyses. The

results of the non-LOCA analyses indicate

that operation with the RSGs in service is

acceptable. As a result of the analyses and

evaluations performed in support of the

RSGs, the ANO specific safety parameters

and regulatory limits are protected.

Therefore, the proposed TS changes will not

significantly increase the probability of an

accident previously analyzed.

Loss of Coolant Accidents (LOCAs) and

non-LOCA safety analyses supporting the

proposed changes have been performed and

have demonstrated conformance with all

applicable Licensing Basis acceptance

criteria. Although calculated radiological

doses using newer, more conservative

methods increase for some non-LOCA events

(requiring a revision to Chapter 15 of the

SAR), the results are within the acceptance

criteria of 10 CFR 100. Therefore, the

proposed changes do not involve a

significant increase in the consequences of an

accident previously evaluated.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

Criterion 2—Does Not Create the Possibility

of a New or Different Kind of Accident From

Any Previously Evaluated

The proposed changes to the ANO–2 TSs

are analytically based and require changing

plant setpoints and procedural limits. No

physical modifications are required as a

result of the proposed changes. The RPS/

ESFAS setpoint changes provide functionally

equivalent protection with the RSGs as the

previous setpoint values provided with the

OSGs. Proposed changes in regard to RCS

flow rate and High Linear Power Trip

setpoints associated with conditions where

MSSVs are inoperable represent appropriate

restrictions that have resulted from the

various analyses performed in support of

RSG installation. The additional 8 hours

provided for reducing the High Linear Power

Level trip setpoints is acceptable due to the

low probability of an event occurring within

this period, based on operating experience

which indicates such a time period is

reasonable to complete the changes, and to

provide consistency with the RSTS [Revised

Standard Technical Specifications].

Therefore, the proposed TS changes will not

create the possibility of a new or different

kind of accident than previously analyzed.

A review of both LOCA and non-LOCA

events was performed which confirms that

existing licensing basis methodologies have

been considered and that a new accident

event has not been created.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3—Does Not Involve a Significant

Reduction in the Margin of Safety

LOCA and non-LOCA safety analyses

supporting the proposed changes have been

performed and have demonstrated

conformance within applicable acceptance

criteria. With the increased size of the RSGs

and the change in design characteristics, the

bases for the setpoints in the ANO–2 TSs are

affected. However, based on the new analyses

and evaluations conducted in support of this

license amendment, the new TS setpoints

provide adequate margin to protect

established safety and regulatory limits.

Although calculated offsite radiological doses

increase slightly for some non-LOCA events

documented in Chapter 15 of the ANO–2

SAR, the increases are not considered to be

significant in that the results remain within

the 10 CFR 100 acceptance criteria.

Therefore, this 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

6406 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

amendment request involves no

significant hazards consideration.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Dated of amendment request: July 29,

1999, as supplemented by letters dated

August 8 and August 24, 1999 (NPF–38–

220).

Description of amendment request:

The proposed change modifies

Technical Specifications (TS) 3.8.1.1

and associated Bases by extending the

Emergency Diesel Generator allowed

outage time from 72 hours to ten days.

Additionally, this proposed change adds

Section 6.16, ‘‘Configuration Risk

Management Program’’ to the

Administrative Controls of the 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 operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: The emergency diesel generators

(EDGs) are backup alternating current power

sources designed to power essential safety

systems in the event of a loss of offsite

power. EDGs are not an accident initiator in

any accident previously evaluated. Therefore,

this change does not involve an increase in

the probability of an accident previously

evaluated.

The EDGs provide backup power to

components that mitigate the consequences

of accidents. The proposed changes to

allowed outage times (AOTs) do not affect

any of the assumptions used in deterministic

safety analyses.

In order to fully evaluate the EDG AOT

extension, probabilistic safety analysis

methods were utilized. The results of these

analyses indicate no significant increase in

the risk of an accident previously evaluated.

These analyses are detailed in CE NPSD–996,

Combustion Engineering Owners Group

‘‘Joint Applications Report for Emergency

Diesel Generators AOT Extension.’’

The Configuration Risk Management

Program is an Administrative Program that

assesses risk based on plant status. Adding

the requirement to implement this program

for Technical Specification 3.8.1.1 ACTION b

does not affect the probability or the

consequences of an accident.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different type of

accident from any accident previously

evaluated?

Response: The proposed change does not

change the design or configuration of the

plant. No new method of plant operation is

involved.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: The proposed changes do not

affect the Technical Specification limiting

conditions for operation or their bases which

support the deterministic analyses used to

establish the margin of safety. Evaluations

used to support the requested Technical

Specification changes have been

demonstrated to be either risk neutral or risk

beneficial depending on precise plant

conditions. These evaluations are detailed in

CE NPSD–996.

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.

Attorney for licensee: N.S. Reynolds,

Esquire, Winston & Strawn 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Dated of amendment request: July 29,

1999, as supplemented by letter dated

August 24, 1999 (NPF–38–221).

Description of amendment request:

The proposed change modifies

Technical Specifications (TS) 3.6.2.1 to

extend the allowable outage time to

seven days for one Containment Spray

System (CSS) train inoperable. A new

ACTION has been added to provide a

shutdown requirement for the

inoperability of two CSSs. Additionally,

the APPLICABILITY is being changed to

provide an end state of MODE 4.

Associated TS Bases changes are

included.

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 operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: The Containment Spray System

(CSS) is part of the Containment

Depressurization and Cooling System.

Inoperable CSS components are not accident

initiators in any accident previously

evaluated. Therefore, this change does not

involve an increase in the probability of any

accident previously evaluated.

The CSS system is primarily designed to

mitigate the consequences of a Loss of

Coolant Accident (LOCA) or Main Steam

Line Break (MSLB). These proposed changes

do not affect any of the assumptions used in

the deterministic LOCA or MSLB analyses.

Hence the consequences of accidents

previously evaluated do not change.

In order to fully evaluate the CSS AOT

[Allowed Outage Time] extension,

probabilistic safety assessment (PSA)

methods were utilized. The results of these

analyses show no significant increase in the

core damage frequency. These analyses are

detailed in report CE NPSD–1045,

‘‘Modifications To The Containment Spray

System, and Low Pressure Safety Injection

System Technical Specifications.’’

The Configuration Risk Management

Program is an Administrative Program that

assesses risk based on plant status. Adding

the requirement to implement this program

for Technical Specification 3.6.2.1 does not

affect the probability or the consequences of

an accident.

Analyzed events are assumed to be

initiated by the failure of plant structures,

systems or components. Allowing an

extended AOT or changing the

APPLICABILITY does not increase the

probability that a failure leading to an

analyzed event will occur. The CSS

components are passive until an actuation

signal is generated. This change does not

increase the failure probability of the CSS

components. As such, the probability of

occurrence for a previously analyzed

accident [is] not significantly increased.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different type of

accident from any accident previously

evaluated?

Response: The proposed change does not

change the design or configuration of the

plant. No new equipment is being

introduced, and installed equipment is not

being operated in a new or different manner.

There is no change being made to the

parameters within which the plant is

operated, and the setpoints at which

protective or mitigative actions are initiated

are unaffected by this change. No alteration

in the procedures which ensure the plant

remains within analyzed limits is being

proposed, and no change is being made to the

procedures relied upon to respond to an offnormal event. As such, no new failure modes

are being introduced. The proposed change

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6407

will only provide the plant some flexibility

in the AOT and chang[es] the

APPLICABILITY. The change does not alter

assumptions made in the safety analysis and

licensing basis. Therefore, the change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: The proposed changes do not

affect the limiting conditions for operation or

their bases used in the deterministic analysis

to establish the margin of safety. PSA

evaluations were used to evaluate these

changes. These evaluations demonstrate that

the changes involve no significant increase in

risk. These evaluations are detailed in report

CE NPSD–1045. The margin of safety is

established through equipment design,

operating parameters, and the setpoints at

which automatic actions are initiated. None

of these are adversely impacted by the

proposed change. Sufficient equipment

remains available to actuate upon demand for

the purpose of mitigating a transient event.

The proposed change, which allows

operation to continue for up to 7 days with

components inoperable in one CSS train, is

acceptable based on the remaining CSS

components providing 100% of the required

CSS flow. The reduced potential for a selfinduced plant transient resulting from unit

shutdown required for a second inoperable

CSS train is minimized. Therefore, the

change does not involve a significant

reduction in the margin of safety, and is

offset by minimizing the potential for a selfinduced plant transient.

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.

Attorney for licensee: N.S. Reynolds,

Esquire, Winston & Strawn 1400 L Street

NW., Washington, DC 20005–3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations Inc., Docket No. 50–382,

Waterford Steam Electric Station, Unit 3, St.

Charles Parish, Louisiana

Date of amendment request: October 18,

1999.

Description of amendment request: The

proposed change modifies Technical

Specification (TS) 3.6.2.2 Limiting Condition

for Operation to allow Waterford Steam

Electric Station, Unit 3 to operate with two

independent trains of containment cooling,

consisting of one cooler per train, operable

during modes 1, 2, 3, and 4. Associated

changes to the TS Bases have been proposed.

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 operation of the facility in

accordance with these proposed changes

involve a significant increase in the

probability or consequence of an accident

previously evaluated?

Response: The proposed change to

Technical Specification (TS) 3.6.2.2 reduces

the number of Containment Fan Coolers

(CFC) from two to one required to be

operable in each train of the Containment

Cooling System for modes 1, 2, 3, and 4. This

change does not create any new system

interactions and has no impact on operation

or function of any system or equipment in a

way that could cause an accident. The CFCs

are not an initiator of any events nor affect

any accident initiators of any events analyzed

in Chapter 15 of the UFSAR [Updated Final

Safety Analysis Report]. Therefore this

change will not impact the probability of

occurrence of an accident.

The results of the reanalysis of the limiting

Loss of Coolant Accident (LOCA) and Main

Steam Line Break (MSLB) accidents show

that the consequences of an accident

previously evaluated are not increased by the

change in the required number of operable

CFCs. The limiting accidents affected by the

proposed changes are identified below:

The peak containment pressure following

the limiting LOCA (Double Ended Hot Leg

Slot Break with minimum safety injection

flow) was determined to be 35.2 psig [pounds

per square inch, gauge] as compared to the

current licensing basis limiting LOCA

(Double Ended Suction Leg Slot Break with

minimum safety injection flow) peak

pressure of 43.1 psig.

The peak containment pressure at 24 hours

following the start of the limiting LOCA

(Double Ended Discharge Leg Slot Break with

minimum safety injection flow) and the

operation of one containment spray train and

one partially flooded CFC operable was

determined to be 15.5 psig with a peak

pressure of 33.27 psig as compared to the

current licensing basis of 14.9 psig with a

peak pressure of 42.9 psig. The current

licensing basis limiting LOCA is the Double

Ended Suction Leg Slot Break with maximum

safety injection flow and the operation of one

containment spray train and two operable

CFCs.

The peak containment pressure following

the limiting MSLB (102% power with failure

of one containment heat removal train

consisting of one containment spray pump

and one CFC operable) was determined to be

42.68 psig as compared to the current

licensing basis peak pressure of 42.9 psig.

The current licensing basis limiting MSLB is

75% power with the failure of one train of

containment heat removal system consisting

of one containment spray train and two

operable CFCs.

The peak containment equipment

qualification temperature following the

limiting MSLB (102% power with the failure

of one MSIV to close) was determined to be

397.4 °F as compared to the current licensing

basis peak temperature of 409.1 °F. The

current licensing basis limiting MSLB is

102% power with two CFCs per train

operable and the failure of one train of

containment spray.

These values above demonstrate that the

containment design basis pressure and

equipment qualification temperature of 44

psig and 413.5 °F, respectively, are not

exceeded and the containment pressure at 24

hours after start of the limiting LOCA is less

than 50% of the peak pressure.

The results of the containment response

analysis discussed above satisfy the

following NRC Staff Standard Review Plan

(SRP) section 6.2.1.1.A guidance document

acceptance criteria for a PWR [Pressurized

Water Reactor] dry containment.

The peak calculated containment pressure

following a Loss of Coolant Accident (LOCA)

or Main Steam Line Break (MSLB) should be

less than the containment design pressure.

To satisfy the requirements of GDC

[General Design Criteria] 38 to rapidly reduce

the containment pressure, the containment

pressure should be reduced to less than 50%

of the peak calculated pressure for the design

basis LOCA within 24 hours after the

postulated accident.

Thus, revising the containment cooling

system TS to require only one operable CFC

per train results in acceptable containment

response and therefore, will not adversely

impact the consequences of accidents

previously evaluated.

Therefore, the proposed changes will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Will the operation of the facility in

accordance with these proposed changes

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response: The proposed change to reduce

the number of operable Containment Fan

Coolers (CFC) from two to one in each train

of the Containment Cooling System for

modes 1, 2, 3, and 4 does not alter the

operation of the CFCs. Although only one of

the two CFCs per train is required to be

operable, the manner in which the CFCs

perform their safety function is not changed.

All four CFCs (two per train) will be

maintained operable to the extent possible to

provide the greatest defense in depth and

operating flexibility.

This proposed change does not involve a

change in plant design, nor does it involve

any potential initiating events that would

create any new or different kind of accident.

This proposed change does not alter the way

in which the plant is operated in a manner

that would create a new or different accident.

Therefore, since no hardware modifications

will be made, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will the operation of the facility in

accordance with these proposed changes

involve a significant reduction in a margin of

safety?

Response: The proposed change revises TS

3.6.2.2, Containment Cooling System. This

change revises the required number of fan

coolers from two fan coolers per train to one

fan cooler per train. As described in the

containment depressurization and cooling

6408 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

system Technical Specification Bases, the

containment cooling system is designed to

maintain the post accident containment peak

pressure below its design value of 44 psig.

The system is also designed to reduce the

containment pressure by a factor of [two]

from its post-accident peak within 24 hours.

The analyses that have been performed to

support this Technical Specification change

have shown that the peak containment

pressure remains below 44 psig, the 24-hour

containment pressure is less than half the

peak pressure, and the containment peak

temperature remains below the maximum

temperature of 413.5 °F provided in the Bases

for Technical Specifications 3.6.2.1 and

3.6.2.2. In comparison of the current safety

margins to the safety margins that would

exist if the proposed changes were in effect,

the results of the analyses, illustrated below,

show an increase in the margin of safety for

containment pressure and equipment

qualification temperature following the

associated limiting LOCA and MSLB.

The peak containment pressure following

the limiting LOCA was determined to be 35.2

psig as compared to the current licensing

basis limiting LOCA peak pressure of 43.1

psig.

The peak containment pressure at 24 hours

following the start of the limiting LOCA was

determined to be 15.5 psig with a peak

pressure of 33.27 psig as compared to the

current licensing basis of 14.9 psig with a

peak pressure of 42.9 psig.

The peak containment pressure following

the limiting MSLB was determined to be

42.68 psig as compared to the current

licensing basis peak pressure of 42.9 psig.

The peak containment equipment

qualification temperature following the

limiting MSLB was determined to be 397.4 °F

as compared to the current licensing basis

peak temperature of 409.1 °F.

This proposed change does not adversely

impact a margin of safety, involve a change

in plant design, or have any affect on the

plant protective barriers. Therefore, the

proposed change will not involve a

significant reduction in the margin of safety.

The Nuclear Regulatory Commission

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.

Attorney for licensee: N. S. Reynolds,

Esquire, Winston & Strawn 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

December 13, 1999.

Description of amendment request:

The licensee proposes to change the

license to delete an expired license

condition and to make some editorial

and administrative changes to correct or

clarify the license.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not adversely

affect accident initiators or precursors nor

alter the design assumptions, conditions, and

configuration of the facility or the manner in

which the plant is operated and maintained.

In addition, the proposed changes do not

affect the manner in which the plant

responds in normal operation, transient or

accident conditions nor do they change any

of the procedures related to operations of the

plant. The proposed changes do not alter or

prevent the ability of structures, systems and

components (SSCs) to perform their intended

function to mitigate the consequences of an

initiating event within the acceptance limits

assumed in the Updated Final Safety

Analysis Report (UFSAR). The proposed

changes are administrative and editorial in

nature and only correct, update and modify

the Operation License.

The proposed changes do not affect the

source term, containment isolation or

radiological release assumptions used in

evaluating the radiological consequences of

an accident previously evaluated in the

Seabrook Station UFSAR. Further, the

proposed changes do not increase the types

and amounts of radioactive effluent that may

be released offsite, nor significantly increase

individual or cumulative occupational/

public radiation exposures.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes are administrative

in nature and only correct, update and clarify

the Seabrook Station Operating License. The

proposed changes do not modify the facility

nor do they modify the manner in which the

plant will be operated nor do they affect the

plant’s response to normal, transient or

accident conditions. The changes do not

introduce a new mode of plant operation.

The plant’s design basis are not revised and

the current safety analyses will remain in

effect and the plant will continue to be

operated in accordance with the existing

Technical Specifications. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes are administrative

and editorial changes to the Seabrook Station

Operating License that do not revise the

Technical Specifications or the bases for the

Technical Specifications. The safety margins

established through Limiting Conditions for

Operation, Limiting Safety System Settings

and Safety Limits as specified in the

Technical Specifications are not revised nor

is the plant design or its method of operation

revised by the proposed changes. Since there

will be no changes to the physical design or

operation of the plant, the proposed changes

do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Section Chief: James W. Clifford.

Omaha Public Power District, Docket No

50–285, For Calhoun Station, Unit No.

1, Washington County, Nebraska.

Date of amendment request: March

18, 1998.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications 2.15(4) and

2.15(5) to identify (1) all indication

functions and control functions required

for the alternate (remote) shutdown

system (alternate shutdown panel and

auxiliary feedwater panel), (2) panel

locations of the functions, and (3) the

number of channels required.

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 proposed changes to Technical

Specifications Section 2.15(4) and 2.15(5)

identify functions, instruments, and controls

along with their location and the number of

required channels. The new Technical

Specifications section addresses the

regulatory requirements for equipment

required for Alternative and Dedicated

Shutdown Capability per 10 CFR Part 50

Appendix R. It will ensure that proper

Limiting Conditions for Operation are

entered for equipment or functional

inoperability. There are no physical

alterations being made to the Alternate

Shutdown Panel and the Auxiliary

Feedwater Panel or related systems.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6409

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes will not result in

any physical alterations to the Alternate

Shutdown Panel or the Auxiliary Feedwater

Panel, or any plant configuration, systems,

equipment, or operational characteristics.

There will be no changes in operating modes,

or safety limits, or instrument limits. With

the proposed changes in place, Technical

Specifications retain requirements for the

Alternate Shutdown Panel and the Auxiliary

Feedwater Panel. Therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from any

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes clarify the

regulatory requirements for the Alternative

and Dedicated Shutdown Capability as

defined by 10 CFR Part 50, Appendix R. The

proposed changes will not alter any physical

or operational characteristics of the Alternate

Shutdown Panel and the Auxiliary

Feedwater Panel and their associated systems

and equipment. Therefore, the proposed

changes do not involve a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Section Chief: Stephen Dembek.

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:

November 29, 1999.

Description of amendment request:

The amendment would adopt the

‘‘Standard Test Method for Nuclear

Grade Activated Carbon’’ for charcoal

filter laboratory testing with certain

exceptions.

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 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

adoption of the new test method and

acceptance criteria of ASTM [American

Society for Testing and Materials] D3803–

1989, with the exceptions as identified in the

Technical Specifications, for activated

charcoal filters does not involve any

modifications to the plant, will not require

changes to how the plant is operated nor will

it affect the operation of the plant. Adoption

of these provisions ensures compliance with

the new test standard of ASTM D 3803–1989.

Adoption of new test method will not cause

an accident and therefore cannot involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(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

accident previously evaluated. The adoption

of the new test method and acceptance

criteria of ASTM D 3803–1989, with the

exceptions as identified in the Technical

Specifications, for activated charcoal filters

does not involve any modifications to the

plant, will not require changes to how the

plant is operated nor will it affect the

operation of the plant. Adoption of new test

method will not cause an accident and

therefore cannot create the possibility of a

new or different kind of accident from any

accident previously evaluated.

(3) Does the proposed license 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 use of

outdated test protocols or inappropriate test

conditions can lead to an overestimation of

the charcoal filters’ ability to adsorb

radioiodine following an accident. The

adoption of the new test method and testing

criteria of ASTM D 3803–1989, with the

exceptions as identified in the Technical

Specifications, for activated charcoal filters

would ensure at least a safety factor of two

is maintained. Thus, the proposed change

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Mr. David E.

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Section Chief: Marsha

Gamberoni, Acting.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request:

November 19, 1999.

Description of amendment request:

The proposed amendment would revise

Sections 4.7.A and 4.11.B of the

Appendix A Technical Specifications

(TSs) to the James A. FitzPatrick

Operating License to adopt the

surveillance test methods and

performance criteria detailed in NRC

Generic Letter 99–02 for laboratory

testing of nuclear-grade charcoal. The

proposed amendment also would

reduce the minimum allowable Standby

Gas Treatment System (SGTS) and

Control Room Emergency Ventilation

Air Ventilation Supply System

(CREVASS) charcoal filter efficiencies

specified in the TSs to those assumed in

the updated radiological dose

calculations.

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 FitzPatrick plant in

accordance with the proposed amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92(c)

since it would not:

(1) Involve an increase in the probability or

consequences of an accident previously

evaluated.

These changes are not modifications to the

plant. They will not require changes to how

the plant is operated, nor will they affect the

operation of the plant.

Changes to these test methods will not

cause an accident and therefore cannot

increase the probability of an accident.

Calculated radiological doses increase as a

result of reductions in assumed charcoal

efficiencies, but remain within regulatory

limits. Radiological doses at the site

boundary and low population zone are less

than 25 percent of 10 CFR 100 criteria. Postaccident doses to control room operators are

also within regulatory limits.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

These changes are not modifications to the

plant, nor will they require changes to how

the plant is operated. Changes to these test

methods will not cause an accident, and

therefore cannot create the possibility of a

new or different kind of accident.

(3) Involve a significant reduction in a

margin of safety.

Evidence has been presented which

contend that activated charcoal testing

performed to test standards other than ASTM

[American Society for Testing and Materials]

D3803–1989 may be inaccurate and may

overestimate its adsorption capabilities. The

adoption of ASTM D3803–1989 charcoal

performance standards and plant-specific test

parameters ensures adequate safety margins.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

6410 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / 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.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Section Chief: Marsha

Gamberoni, Acting.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request:

December 20, 1999.

Description of amendment request:

This application for an amendment to

the James A. FitzPatrick Technical

Specifications (TS) proposes a change to

the Main Steam Isolation Valve (MSIVs)

closure scram setpoint. The proposed

amendment changes the MSIV closure

scram Trip Level Setting from ≤10% to

≤15% valve closure.

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 will not

significantly increase the probability or

consequences of any previously evaluated

accidents.

This proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. This proposed change to the MSIV

scram trip setpoint from 90% open to 85%

open results in the MSIV closure scram event

having a slight delay in the initiation of the

reactor scram. The MSIV scram event is not

an accident initiator and will not increase the

probability or consequence of any accident

previously evaluated. An evaluation of

events concluded that the MSIV direct scram

remains a relatively low consequence event

and has no effect on any operating limits.

The limiting event analyzed for the reactor

vessel overpressure event in the James A.

FitzPatrick Updated Final Safety Analysis

Report (UFSAR) is the MSIV closure

terminated by a high neutron flux scram,

which does not take credit for the MSIV

closure valve position scram. In addition, an

evaluation of the Main Steam Line Break

outside containment event concludes that

there is no impact on PCT [Peak Clad

Temperature] or break flow and that the

results presented in the UFSAR are

bounding. An evaluation of the impact on

containment was also made. The

containment response is evaluated for much

more severe events such as a Loss of Coolant

Accident (LOCA) or stuck open Safety Relief

Valve (SRV), thus the change in MSIV scram

setpoint has no impact on the containment

analysis. Therefore, changing the MSIV

closure scram setpoint from 90% open to

85% open does not significantly increase the

probability or consequences of any

previously evaluated accidents.

2. The proposed change will not create the

possibility of a new or different kind of

accident.

This proposed change to the MSIV scram

trip setpoint from 90% open to 85% open

results in the MSIV closure scram event

having a slight delay in the initiation of the

reactor scram, and does not introduce a new

or different kind of accident previously

analyzed. An evaluation of the event

determined that the MSIV closure scram

remains a relatively low consequence event

and has no effect on any operating limits.

The limiting event analyzed for the reactor

vessel overpressure event in the UFSAR is

the MSIV closure terminated by a high

neutron flux scram, which does not take

credit for the MSIV closure valve position

scram. The proposed change will not create

the possibility of a new or different kind of

accident.

3. The proposed change will not involve a

significant reduction in a margin of safety.

This proposed change to the MSIV scram

trip setpoint from 90% open to 85% open

results in the MSIV closure scram event

having a slight delay in the initiation of the

reactor scram. An evaluation of the event

determined that the MSIV closure scram

remains a relatively low consequence event

and has no effect on any operating limits. In

addition, an evaluation of the Main Steam

Line Break outside containment event

concludes that there is no impact on PCT or

break flow and that the results presented in

the UFSAR are bounding. An evaluation of

the impact on containment was also made.

The containment response is evaluated for

much more severe events such as a LOCA or

stuck open SRV, thus the change in MSIV

scram setpoint has no impact on the

containment analysis. Changing the MSIV

valve position scram setpoint from ≤10% to

≤15% of valve closure will allow the limit

switches to be positioned such that both

scram and indicating limit switches can be

coordinated and provide accurate and

reliable valve position indication in the

control room. Therefore, changing the MSIV

closure scram setpoint from 90% open to

85% open does not involve a significant

reduction in a margin of safety and provides

a net benefit to plant operations.

The proposed change will not increase the

probability or consequences of any

previously analyzed accident, introduce any

new or different kind of accident previously

evaluated, or significantly reduce existing

margin to safety. Therefore, the proposed

license amendment will not involve a

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Section Chief: Marsha

Gamberoni, Acting.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia.

Date of amendment request: August

30, 1999.

Description of amendment request:

The proposed amendments would

revise Technical Specification 5.2.2 in

order to raise the level of the approval

authority for deviations from the

guidelines provided to minimize unit

staff overtime.

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. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed license amendment would

strengthen the administrative controls that

permit plant personnel to work beyond those

limits outlined in the TS’s. As a result, there

will be greater scrutiny on the amount of

overtime being utilized to perform safetyrelated function. Therefore, it has been

determined that operation of the facility in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

There will be no physical changes to the

systems, components or structure of the

facility as a result of this proposed license

amendment. The initial assumptions of the

design accident analyses will be unaffected.

Therefore, operation of the facility in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

This amendment raises the administrative

level of management approval required for

overtime in excess of the limits outlined in

the TS. Therefore, operation of the facility in

accordance with the proposed amendment

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.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6411

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308–2216.

NRC Section Chief: Richard L. Emch,

Jr.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of amendment request: October

4, 1999.

Description of amendment request:

The proposed amendments would

revise Technical Specification 5.5.6,

‘‘Prestressed Concrete Containment

Tendon Surveillance Program,’’ to

incorporate three exceptions to

Regulatory Guide (RG) 1.35, Revision 2,

1976. The exceptions concern the

number of tendons detensioned,

inspection of concrete adjacent to

vertical tendons, and the time during

which areas adjacent to tendons are

inspected.

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 proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

No. The proposed change only clarifies TS

requirements for the containment tendon

surveillance program. The proposed

clarification has been previously reviewed

and approved by the NRC staff with

Amendments 23 and 4, and is consistent

with current regulatory guidance. As such,

the proposed change is essentially

administrative in nature. The containment

tendon surveillance program has no impact

on the probability of any accident initiators,

and it will continue to ensure containment

structural integrity. Therefore, the proposed

change does not involve a significant

increase in the consequences of an accident

previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any previously evaluated?

No. The proposed change only clarifies TS

requirements for the containment tendon

surveillance program. The proposed

clarification has been previously reviewed

and approved by the NRC staff with

Amendments 23 and 4, and is consistent

with current regulatory guidance. As such,

the proposed change is essentially

administrative in nature. Plant design and

operation will not be changed, and no other

safety related or important to safety

equipment is affected by the proposed

change. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety.

No. The proposed change only clarifies TS

requirements for the containment tendon

surveillance program. The proposed

clarification has been previously reviewed

and approved by the NRC staff with

Amendments 23 and 4, and is consistent

with current regulatory guidance. As such,

the proposed change is essentially

administrative in nature. The containment

prestressing system will continue to perform

its function to ensure containment structural

integrity. Therefore, the proposed 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.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308–2216.

NRC Section Chief: Richard L. Emch,

Jr.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: June 25,

1999 (TS 98–016).

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) and

TS Bases to reflect application of the

Westinghouse generic Best Estimate

Large Break Loss-of-Coolant Accident

Analysis methodology using the

WCOBRA/TRAC computer code.

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:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes involve use of the

Best Estimate Large Break loss-of-coolant

accident (LOCA) analysis methodology and

associated technical specification changes.

Accumulator water level set points will be

revised from [greater than or equal to] 7717

gallons and [less than or equal to] 8004

gallons to [greater than or equal to] 7630

gallons and [less than or equal to] 8000

gallons to provide the plant with an

increased operating range. The plant

conditions assumed in the analysis,

including the accumulator water level

instrumentation changes, are bounded by the

design conditions for all equipment in the

plant.

Therefore, there will be no increase in the

probability of a LOCA. The consequences of

a LOCA are not being increased, since it is

shown that the emergency core cooling

system (ECCS) is designed so that its

calculated cooling performance conforms to

the criteria contained in 10 CFR 50.46,

Paragraph b. The small break LOCA analysis

assumes only a nominal accumulator water

level which is the same nominal value

assumed in this analysis, therefore, the small

break LOCA analysis is unaffected by the

increase in the accumulator range. Also, the

increased safety analysis range in

accumulator water volume (+/¥15 cubic

feet) has an insignificant effect on the

containment related analyses.

The post-LOCA containment sump boron

calculation assumes a minimum accumulator

volume which bounds (is smaller than) the

1005 cubic feet (7518 gallons) value

supported by the Best Estimate Large Break

LOCA analysis. Also, the hot leg switchover

calculation models a maximum accumulator

volume which is not bounded by the 1095

cubic feet (8191 gallons) maximum value

supported by the Best Estimate Large Break

LOCA analysis. However, an evaluation

concludes that the Watts Bar hot leg

switchover time is unaffected by the

difference in maximum volumes.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

No new modes of plant operation are being

introduced by the new analysis or by the

changes in instrumentation setpoints for

accumulator water level. The parameters

assumed in the analysis are within the design

limits of existing plant equipment. Therefore,

the proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

It has been shown that the analytic

technique used in the analysis realistically

describes the expected behavior of the WBN

Unit 1 reactor system during a postulated

loss of coolant accident. Uncertainties have

been accounted for as required by 10 CFR

50.46. The physical setpoint changes to

accumulator water level instrumentation are

bounded by the uncertainty evaluation

addressing accumulator water level. A

sufficient number of loss of coolant accidents

with different break sizes, different locations,

and other variations in properties have been

considered to provide assurance that the

most severe postulated LOCAs were

evaluated. It has been shown by the analysis

that there is a high level of probability that

all criteria contained in 10 CFR 50.46,

Paragraph b are met.

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.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

6412 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

400 West Summit Hill Drive, ET l0H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

December 14, 1999.

Description of amendment request:

This proposed change would revise the

Vermont Yankee (VY) Technical

Specifications (TS) by relocating the

procedural details of the Radiological

Effluent Technical Specifications

(RETS) to the Offsite Dose Calculation

Manual (ODCM). The TS would also be

revised to relocate procedural details

associated with solid radioactive wastes

to the Process Control Program (PCP). In

addition, the TS definition for

‘‘solidification’’ would be relocated to

the VY Technical Requirements Manual

(TRM).

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 operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not affect

accident initiators or precursors and do not

alter the design assumptions, conditions,

configuration of the facility, or the manner in

which the plant is operated. The proposed

changes do not alter or prevent the ability of

structures, systems, or components to

perform their intended safety function to

mitigate the consequences of an initiating

event within the acceptance limits assumed

in the Updated Final Safety Analysis Report

(UFSAR). The proposed changes are

administrative in nature and do not change

the level of programmatic controls and

procedural details relative to radiological

effluents.

Implementation of programmatic controls

for RETS already in TS will assure that the

applicable regulatory requirements

pertaining to the control of radioactive

effluents will continue to be maintained.

Since there are no changes to previous

accident analysis, the radiological

consequences associated with these analyses

remain unchanged.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not alter the

design assumptions, conditions,

configuration of the facility, or the manner in

which the plant is operated. The proposed

changes have no impact on component or

system interactions. The proposed changes

are administrative in nature and do not

change the level of programmatic controls

and procedural details relative to radiological

effluents.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated for Vermont Yankee.

3. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant reduction in a margin of safety.

There is no impact on equipment design or

operation, and there are no changes being

made to the TS-required safety limits or

safety system settings that would adversely

affect plant safety as a result of the proposed

changes. The proposed changes are

administrative in nature and do not change

the level of programmatic controls and

procedural details relative to radiological

effluents. A comparable level of

administrative controls will continue to be

applied to those specifications being

relocated to the ODCM, PCP, or TRM.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

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: October

28, 1999, as supplemented December

21, 1999.

Description of amendment request:

The proposed changes will remove the

operability and surveillance

requirements of Technical

Specifications (TS) Section 3/4.6.4.3,

‘‘Waste Gas Charcoal Filter System’’

from the TS and relocate them to the

Technical Requirements Manual.

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:

A waste gas decay tank rupture is highly

unlikely, as the waste gas decay tanks are

designed and constructed to stringent quality

control standards, are provided with pressure

relief valves to prevent overpressurization,

are missile-shielded by installation below

grade, and have their gaseous contents

controlled to prevent potentially explosive

mixtures. The entire gaseous content of the

waste gas decay tank is assumed to be

released to the atmosphere as a ground-level

release * * *. The waste gas charcoal filter

system is not credited for any mitigation of

the release in the accident analysis for a

waste gas decay tank rupture. In addition, the

releases associated with a waste gas decay

tank rupture are bounded by the existing

LOCA [loss-of-coolant accident] releases.

Specifically, operation of the North Anna

Power Station in accordance with the

proposed Technical Specification changes

will not:

[1.] Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Relocating the operability and surveillance

requirements for The Waste Gas Charcoal

System to the TRM [Technical Requirements

Manual] does not change the operation of the

plant. The plant and the radioactive gas

waste system will not be operated differently.

No new accident initiators are established as

a result of the proposed changes. Therefore,

the probability of occurrence is not increased

for any accident previously evaluated.

Relocating the operability and surveillance

requirements for The Waste Gas Charcoal

Filter to the TRM does not [a]ffect the

gaseous releases to the environment, which

are controlled by the ODCM [Offsite Dose

Calculation Manual]. Additionally, no credit

for these filters is taken in the accident

analysis for Waste Gas Decay Tank rupture.

Therefore, there is no increase in the

consequences of any accident previously

analyzed.

[2.] Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not affect the

operation of the plant. The gaseous waste

systems will not be operated differently as a

result of the proposed changes. No new

accident or event initiators are created

moving the operability and surveillance

requirements for The Waste Gas Charcoal

Filter to the TRM. Therefore, the proposed

changes do not create the possibility of any

accident or malfunction of a different type.

[3.] Involve a significant reduction in the

margin of safety as defined in the bases on

any Technical Specifications.

The proposed changes have no effect on

any safety analyses assumptions. The waste

gas charcoal filters are not used to mitigate

the consequence[s] [of] a Waste Gas Decay

Tank rupture. The accident analysis assumes

total release of the radioactiv[ity] in the

Waste Gas Decay Tank in the accident

analysis. Therefore, the proposed changes do

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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6413

involves no significant hazards

consideration.

Attorney for licensee: Mr. Donald P.

Irwin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Section Chief: Richard L. Emch

Jr.

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

Dated of amendment request:

November 29, 1999.

Description of amendment request:

The proposed changes will modify the

Technical Specifications in Section

4.7.7.1 for the Control Room Emergency

Habitability System and Section 4.7.8.1

for the Safeguards Area Ventilation

System. The changes will require

laboratory testing of the charcoal filter

carbon to be consistent with American

Society for Testing and Materials

(ASTM) Standard D3803–1989.

Basis for proposed no significant

hazards consideration determination: In

10 CFR 50.92 three criteria are provided

to determine whether a proposed

license amendment involves a

significant hazards consideration. No

significant hazards consideration is

involved if operation of the facility 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. 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:

Virginia Electric and Power Company has

reviewed the requirements of 10 CFR 50.92

as they relate to the proposed changes for the

North Anna Units 1 and 2 and determined

that a significant hazards consideration is not

involved. The proposed Technical

Specification changes adopt the nucleargrade charcoal testing requirements of ASTM

D3803–1989 and do not affect the design or

operation of the plant. The changes also do

not involve any physical modification to the

plant or result in a change in a method of

system operation. The adoption of the 1989

edition of ASTM D3803 provides assurance

that testing of nuclear-grade activated

charcoal of ventilation systems is being

performed with a suitable standard to ensure

that charcoal adsorbers are capable of

performing their required safety function and

that the regulatory requirements regarding

onsite and offsite dose consequences

continue to be satisfied. The changes do not

create an unreviewed safety question.

(a) The proposed changes modify

surveillance testing requirements and do not

affect plant systems or operation and

therefore do not increase the probability or

the consequences of an accident previously

evaluated. The proposed surveillance

requirements adopt ASTM D3803–1989 as

the laboratory method for testing samples of

the charcoal adsorber in response to NRC’s

Generic Letter 99–02. This method of testing

charcoal adsorbers has been approved by the

NRC as an acceptable method for

determining methyl iodide removal

efficiency. Since the charcoal adsorbers are

used to mitigate the consequences of an

accident, the more accurate the test, the

better assurance we have that we remain

within our accident analysis assumptions.

The laboratory test acceptance criteria

contain a safety factor to ensure that the

efficiency assumed in the accident analysis is

still valid at the end of the operating cycle.

There is no change in the method of plant

operation, system performance or system

design. (b) The proposed changes do not

create the possibility of an accident or

malfunction of a different type than any

evaluated previously.

The proposed changes modify surveillance

testing requirements and do not impact plant

systems or operations and therefore do not

create the possibility of an accident or

malfunction of a different type than

evaluated previously. The proposed

surveillance requirements adopt ASTM

D3803–1989 as the laboratory method for

testing samples of the charcoal adsorber. This

change is in response to NRC’s request in

response to their Generic Letter 99–02. There

is no change in the method of plant operation

or system design. There are no new or

different accident scenarios, transient

precursors, nor failure mechanisms that will

be introduced.

(c) The proposed changes modify

surveillance test requirements and do not

impact plant systems or operations and

therefore do not significantly reduce the

margin of safety. The revised surveillance

requirements adopt ASTM D3803–1989 as

the laboratory method for testing samples of

the charcoal adsorber. The 1989 edition of

this standard imposes very stringent

requirements for establishing the capability

of new and used activated carbon to remove

radio-labeled methyl iodide from air and gas

streams. The results of this test provide a

more conservative estimate of the

performance of nuclear-graded activated

carbon used in all nuclear power plant HVAC

[heating, ventilation, and air conditioning]

systems for the removal of radioiodine. The

laboratory test acceptance criteria contain a

safety factor to ensure that the efficiency

assumed in the accident analysis is still valid

at the end of the operating cycle.

This analysis demonstrates that the

proposed amendment to The North Anna

Units 1 and 2 Technical Specifications does

not involve a significant increase in the

probability or consequences of a previously

evaluated accident, does not create the

possibility of a new or different kind of

accident and 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Mr. Donald P.

Irwin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Section Chief: Richard L. Emch

Jr.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed no Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Dated of amendment request: April

11, 1996, as supplemented April 6,

1998, March 22 and July 29, 1999 (PCN–

460).

Brief description of amendment

request: The proposed amendments

would revise the San Onofre Units 2

and 3 Technical Specification (TS)

related to the containment isolation

valves. Specifically, the licensee

proposed a revision to TS 3.6.3 to

extend the completion times for Section

D.1 and D.2 valves from 4 hours to the

applicable limiting condition for

operation time pertaining to the

engineered safety feature system in

which the valve is installed.

Dated of publication of individual

notice in Federal Register: January 19,

2000 (65 FR 2993), as corrected January

26, 2000 (65 FR 4265).

Expiration date of individual notice:

February 18, 2000.

6414 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

Southern California Edison Company, et

al., Docket Nos.50–361 and 50–362, San

Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of amendment request: January

2, 1998, as supplemented December 13,

1999 (PCN–482).

Brief description of amendment

request: The proposed amendments

would revise the San Onofre Units 2

and 3 Technical Specification (TS)

relating to the Auxiliary Feedwater

(AFW) System. Specifically, the licensee

proposed to revise TS 3.7.5 to add a

note that states: ‘‘The steam driven AFW

pump is OPERABLE when running and

controlled manually to support plant

start-ups, plant shut-downs, and AFW

pump and valve testing.’’

Date of publication of individual

notice in Federal Register: January 19,

2000 (65 FR 2991), as corrected January

26, 2000 (65 FR 4265).

Expiration date of individual notice:

February 18, 2000.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of amendment request: January

11, 1999, as supplemented November

29, 1999 (PCN–499).

Brief description of amendment

request: The proposed amendments

would revise the San Onofre Units 2

and 3 Technical Specification (TS)

3.7.6, ‘‘Condensate Storage Tank (CST

T–121 and T–120)’’ to change the

minimum inventory of water

maintained in the condensate storage

tank (T–120) from 280,000 gallons to

360,000 gallons during plant operation

Modes 1, 2 and 3.

Date of publication of individual

notice in Federal Register: January 18,

2000 (65 FR 2648).

Expiration date of individual notice:

February 17, 2000.

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

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room).

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

September 28, 1999.

Brief description of amendment: The

amendment revises Technical

Specification Surveillance Requirement

3.7.6.2 ‘‘Component Cooling Water

(CCW) System’’ to change the CCW

pump automatic start actuation signal

basis from Engineered Safety Feature

Actuation Signal to Loss-of-Power

Diesel Generator.

Date of issuance: January 21, 2000.

Effective date: January 21, 2000.

Amendment No.: 186.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR

59798).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 21,

2000.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request: May 11,

1999, as supplemented June 3 and July

28, 1999.

Brief description of amendment: This

amendment deletes from the Defueled

Technical Specifications (DTS)

subsection 1.16, ‘‘SITE BOUNDARY,’’

Figure 5.1–1, the Big Rock Point (BRP)

Site Map, and DTS 5.1.1 paragraph

numbering and removes certain sitespecific information from DTS 5.1,

which describes the BRP site. The

amendment also makes editorial

changes to DTSs 6.6.2.5. g, h, and j, and

6.6.2.6.b. because of the changes

associated with DTSs 1.16 and 5.1 and

Figure 5.1–1 described above. Most of

the information removed or deleted

from the DTSs can be found in the BRP

Final Hazards Summary Report.

Date of Issuance: January 13, 2000.

Effective Date: January 13, 2000, to be

implemented within 45 days from date

of issuance. Implementation includes

incorporation of the site boundary

information, as discussed in the staff’s

safety evaluation enclosed with this

amendment, into the next Final Safety

Analysis Report (i.e. the updated Final

Hazards Summary Report for the Big

Rock Point Nuclear Plant) update in

accordance with the schedule in 10 CFR

50.71(e).

Amendment No.: 121.

Facility Operating License No. DPR–6.

The amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: June 16, 1999 (64 FR 32288).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 13,

2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

April 5, 1999 as supplemented by letters

dated May 27, July 6, October 7, and

November 22, 1999.

Brief description of amendments: The

amendments revised the Technical

Specifications to incorporate Topical

Report DPC–NE–3005–P, ‘‘Thermal

Hydraulic Transient Analysis

Methodology.’’

Date of Issuance: January 18, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6415

Amendment Nos.: Unit 1–309; Unit

2–309; Unit 3–309.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 30, 1999 (64 FR 35202),

November 3, 1999 (64 FR 59801).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 18,

2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina.

Date of application of amendments:

September 29, 1999.

Brief description of amendments: The

amendments revise the Containment

Inservice Inspection Program Technical

Specifications related to the

containment leakage testing program

and the pre-stressed concrete

containment tendon surveillance

program.

Date of Issuance: January 18, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: Unit 1—310; Unit

2—310; Unit 3—310.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 17, 1999 (64 FR

62707).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 18,

2000.

No significant hazards consideration

comments received: No.

Entergy Gulf States, Inc., and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request:

December 16, 1999.

Brief description of amendment: The

license amendment revises the River

Bend Station Technical Requirements

Manual, Section TR 3.9.14, and adds a

temporary exception to the current

prohibition for travel of loads in excess

of 1200 pounds over fuel assemblies in

the spent fuel storage pool. The

exception allows the licensee to move

the spent fuel pool (SFP) watertight

gates, which separate the SFP from the

cask and lower transfer pools, in order

to perform repairs on the gates and

watertight seals prior to the end of

Refueling Outage 9. Updated Safety

Analysis Report (USAR) Sections

9.1.2.2.2 and 9.1.2.3.3 are also changed

to reflect the proposed exception.

Date of issuance: January 13, 2000.

Effective date: The license

amendment is effective as of its date of

issuance and shall be implemented in

the next periodic update to the USAR

and TRM in accordance with 10 CFR

50.71(e). Implementation of the

amendment is the incorporation into the

USAR and TRM update, the changes to

the description of the facility as

described in the licensee’s application

dated December 16, 1999, as

supplemented by letters dated

December 21, 1999, and January 10,

2000, and evaluated in the staff’s Safety

Evaluation attached to this amendment.

Amendment No.: 108.

Facility Operating License No. NPF–

47: The amendment revises the

Technical Requirements Manual and

Updated Safety Analysis Report.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (64 FR 71511

dated December 21, 1999). The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by January 20, 2000,

but indicated that if the Commission

made a final NSHC determination, any

such hearing would take place after

issuance of the amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final NSHC

determination are contained in a Safety

Evaluation dated January 13, 2000.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 13,

2000.

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC

20005.

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1,

Pope County, Arkansas

Date of amendment request:

December 16, 1999

Brief description of amendment: The

proposed change would amend

Technical Specification 4.18.5.b to

allow tube 110/60 to remain in service

through the current operating cycle

(cycle 16) with two axial indications

that have potential through-wall depths

greater than the plugging limit. The

axial indications are located in the roll

transition region and are contained

within the upper tubesheet.

Date of issuance: January 13, 2000.

Effective date: As of the date of

issuance.

Amendment No.: 203.

Facility Operating License No. DPR–

51: Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (64 FR 73080

dated December 29, 1999). The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by January 28, 2000,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final NSHC

determination are contained in a Safety

Evaluation dated January 13, 2000.

Attorney for Licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

Entergy Operations, Inc., System Energy

Resources, Inc., South Mississippi

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:

July 20, 1998, as supplemented by letter

dated June 29, 1999.

Brief description of amendment: The

amendment incorporates the Technical

Specification changes necessary for

implementation of the Boiling Water

Reactor Owners’ Group Reactor Stability

Long-Term Solution, Enhanced Option

1–A.

Date of issuance: January 19, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 120 days of issuance.

Amendment No: 141.

Facility Operating License No. NPF–

29: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR

46432).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 19,

2000.

No significant hazards consideration

comments received: No.

6416 Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

October 12, 1999.

Brief description of amendments: The

amendments would revise the Technical

Specifications (TSs) Surveillance

Requirement 4.6.2.2.d for the spray

additive system to relocate the details

associated with the acceptance criteria

and test parameters to the associated

TSs Bases. Additionally, certain

administrative text format changes were

made.

Date of issuance: January 19, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 240 and 221.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR

59804).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 19,

2000.

No significant hazards consideration

comments received: No.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

September 11, 1998, as supplemented

by letters dated January 14, and August

5, 1999.

Brief description of amendments: The

amendments revise the combined

Technical Specifications (TS) for the

Diablo Canyon Power Plant, Unit Nos. 1

and 2 to revise TS 6.8.4f., ‘‘Containment

Polar and Turbine Building Cranes,’’ to

control the operation of the containment

polar cranes in jet impingement zones

during Modes 1, 2, 3, and 4.

Date of issuance: January 12, 2000.

Effective date: January 12, 2000.

Amendment Nos.: Unit 1–137; Unit

2–137.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 21, 1999 (64 FR 19561).

The August 5, 1999, supplemental

letter provided additional clarifying

information, did not expand the scope

of the application as originally noticed,

and did not change the staff’s original

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 12,

2000.

No significant hazards consideration

comments received: No.

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:

September 29, 1999, as supplemented

December 7, 1999.

Brief description of amendment: This

amendment revises the Technical

Specifications to allow, on a one-time

basis only, the Power Authority of the

State of New York to extend the allowed

out-of-service time for the Residual Heat

Removal Service Water (RHRSW)

System from 7 days to 11 days. This

amendment is only applicable during

installation of Modification 99–095 to

the ‘‘A’’ RHRSW Strainer.

Date of issuance: January 28, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 259.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 20, 1999 (64 FR

56532).

The December 7, 1999, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 28,

2000.

No significant hazards consideration

comments received: No.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

August 25, 1999.

Brief description of amendments: The

amendments authorize the licensee to

perform single-cell charging of operable

safety-related batteries by using nonClass 1E single-cell battery chargers,

with proper electrical isolation. The

single-cell chargers would be used to

restore individual cell float voltage to

the normal TS limit.

Date of issuance: January 24, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days.

Amendment Nos.: 226 and 207.

Facility Operating License Nos. DPR–

70 and DPR–75.: Amendments revised

the licenses.

Date of initial notice in Federal

Register: September 22, 1999 (64 FR

51349).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 24,

2000.

No significant hazards consideration

comments received: No.

Rochester Gas and Electric Corporation,

Docket No. 50–244, R. E. Ginna Nuclear

Power Plant, Wayne County, New York

Date of application for amendment:

June 28, 1999.

Brief description of amendment: The

amendment revises the Ginna Station

Improved Technical Specifications

associated with the Reactor Coolant

System Leakage Detection

Instrumentation.

Date of issuance: January 19, 2000.

Effective date: January 19, 2000.

Amendment No.: 76.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 11, 1999 (64 FR 43778)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 19,

2000.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

April 19, 1999, as supplemented by

letter dated November 1, 1999.

Brief description of amendments: The

amendments revised the Technical

Specifications Surveillance

Requirement (SR) 3.3.5.2 and associated

Bases to allow the loss of voltage and

degraded voltage trip setpoints to be

treated as ‘‘nominal’’ values.

Date of issuance: January 19, 2000

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 111—Unit 1; 89—

Unit 2.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 1, 1999 (64 FR

67340).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 19,

2000.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 65, No. 27 / Wednesday, February 9, 2000 / Notices 6417

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request:

September 8, 1999, as supplemented

November 9, 1999.

Brief description of amendments: The

amendments revised Technical

Specification 3/4.8.1, ‘‘A.C. Sources,

Operating,’’ and associated Bases, by

deleting the 18-month surveillance to

subject the standby diesel generator to

inspections in accordance with

procedures prepared in conjunction

with its manufacturer’s

recommendations. The surveillance

requirements have been relocated to the

Technical Requirements Manual.

Date of issuance: January 14, 2000.

Effective date: January 14, 2000, to be

implemented within 30 days.

Amendment Nos.: Unit 1—121 ; Unit

2—109

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 1, 1999 (64 FR

67341).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 14,

2000.

No significant hazards consideration

comments received: No.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request:

September 8, 1999, as supplemented by

letter dated November 9, 1999.

Brief description of amendments: The

amendments revised Technical

Specification (TS) 3/4.8.1, ‘‘A.C.

Sources, Operating,’’ and associated

Bases, by eliminating the requirement

for accelerated testing of the standby

diesel generators and the associated

reporting requirements. The TS Index

was also revised to reflect these

changes.

Date of issuance: January 14, 1999.

Effective date: January 14, 1999.

Amendment Nos.: Unit 1—122 ; Unit

2—110.

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR

59806).

The November 9, 1999, supplement

provided additional clarifying

information that was within the scope of

the original application and Federal

Register notice and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 14,

2000.

No significant hazards consideration

comments received: No

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of application for amendment:

October 21, 1999.

Brief description of amendment: The

amendment corrects two textual errors

and changes the designation of a

referenced figure.

Date of Issuance: January 11, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 183.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 17, 1999 (64 FR

62717).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated January 11,

2000.

For the Nuclear Regulatory Commission.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland, this 2nd day

of February 2000.

For the Nulcear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–2835 Filed 2–8–00; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Proposed Collections; Comment

Request

Upon Written Request; Copies Available

From: Securities and Exchange Commission,

Office of Filings and Information Services,

Washington, DC 20549.

Extension:

Form S–2, SEC File No. 270–60, OMB

Control No. 3235–0072

Form F–1, SEC File No. 270–249, OMB

Control No. 3235–0258

Form F–2, SEC File No. 270–250, OMB

Control No. 3235–0257

Form F–3, SEC File No. 270–251, OMB

Control No. 3235–0256

Form F–7, SEC File No. 270–331, OMB

Control No. 3235–0383

Form F–8, SEC File No. 270–332, OMB

Control No. 3235–0378

Form F–X, SEC File No. 270–336, OMB

Control No. 3235–0379

Form 10–SB, SEC File No. 270–367 OMB

Control No. 3235–0419

Form DF, SEC File No. 270–430, OMB

Control No. 3235–0482

Form T–1, SEC File No. 270–121, OMB

Control No. 3235–0110

Form T–2, SEC File No. 270–122, OMB

Control No. 3235–0111

Form T–3, SEC File No. 270–123, OMB

Control No. 3235–0105

Form T–4, SEC File No. 270–124, OMB

Control No. 3235–0107

Schedule 13E–4F, SEC File No. 270–340,

OMB Control No. 3235–0375

Schedule 14D–1F, SEC File No. 270–338,

OMB Control No. 3235–0376

Schedule 14D–9F, SEC File No. 270–339,

OMB Control No. 3235–0382

Rule 14f–1, SEC File No. 270–127, OMB

Control No. 3235–0108

Rule 12d1–3, SEC File No. 270–116, OMB

Control No. 3235–0109

Notice is hereby given that pursuant

to the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.) the Securities

and Exchange Commission

(Commission) is soliciting comments on

the collections of information

summarized below. The Commission

plans to submit these existing

collections of information to the Office

of Management and Budget for

approval.

Form S–2 is used for registration of

securities of certain issuers. The Form

S–2 provides investors with the

necessary information to make

investment decisions regarding

securities offered to the public. The

likely respondents will be public

companies. The information collected

must be filed with the Commission. All

information is provided to the public

upon request. Form S–2 takes 470

burden hours to prepare and is filed by

101 respondents for a total of 47,470

burden hours.

Form F–1 is a registration statement of

securities of certain foreign private

issuers. Form F–1 provides the public

with the necessary information to make

informed investment decisions

regarding securities offered to the public

by foreign private issuers. The

information provided on Form F–1 is

mandatory. All information on Form F–

1 is reported to the public upon request.

Form F–1 takes approximately 1,868

burden hours to prepare and is filed by

170 respondents. It is estimated that

25% of the 317,560 total burden hours

(79,390 hours) would be prepared by the

company.

Form F–2 is a registration statement of

securities of certain foreign private

issuers. Form F–2 provides the public

with the necessary information to make