37420 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

inspection at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC.

Publically available records will be

accessible electronically from the

ADAMS Public Library component on

the NRC Web site, http://www.nrc.gov

(the Electronic Reading Room).

Dated at Rockville, Maryland, this 6th day

of June 2000.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–15002 Filed 6–13–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meetings

AGENCY HOLDING THE MEETING: Nuclear

Regulatory Commission.

DATE: Weeks of June 12, 19, 26, July 3,

10, and 17, 2000.

PLACE: Commissioner’s Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and closed.

MATTERS TO BE CONSIDERED: 

Week of June 12

Tuesday, June 13, 2000

9:25 a.m.

Affirmation Session (Public Meeting)

a: Final Rule—Clarification of

Regulations to Explicitly Limit

Which Types of Applications Must

Include Antitrust Information

9:30 a.m.

Meeting with Organization of

Agreement States (OAS) and

Conference of Radiation Control

Program Directors (CRCPD) (Public

Meeting) (Contact: Paul Lohaus,

301–415–3340)

1:00 p.m.

Meeting with Korean Peninsula

Energy Development Organization

(KEDO) and State Department

(Public Meeting) (Contact: Donna

Chaney, 301–415–2644)

Week of June 19—Tentative

Tuesday, June 20, 2000

9:25 a.m.

Affirmation Session (Public Meeting)

(If needed)

9:30 a.m.

Briefing on Final Rule—Part 70—

Regulating Fuel Cycle Facilities

(Public Meeting) (Contact: Theodore

Sherr, 301–415–7218)

1:30 p.m.

Briefing on Risk-Informed Part 50,

Option 3 (Public Meeting) (Contact:

Mary Drouin, 301–415–6675)

Wednesday, June 21, 2000

10:30 a.m.

All Employees Meeting (Public

Meeting) (‘‘The Green’’ Plaza Area)

1:30 p.m.

All Employees Meeting (Public

Meeting) (‘‘The Green’’ Plaza Area)

Week of June 26—Tentative

There are no meetings scheduled for

the Week of June 26.

Week of July 3—Tentative

There are no meetings scheduled for

the Week of July 3.

Week of July 10—Tentative

Tuesday, July 11

9:25 a.m.

Afirmation Session (Public Meeting)

(If necessary.)

Week of July 17—Tentative

There are no meetings scheduled for

the Week of July 17.

*The schedule for Commission

meetings is subject to change on short

notice. To verify the status of meetings

call (recording)—(301) 415–1292.

CONTACT PERSON FOR MORE INFORMATION:

Bill Hill (301) 415–1661.

The NRC Commission Meeting

Schedule can be found on the Internet

at: http://www.nrc.gov/SECY/smj/

schedule.htm

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661). In addition, distribution of

this meeting notice over the Internet

system is available. if you are interested

in receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

Dated: June 9, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–15159 Filed 6–12–00; 1:31 pm]

BILLING CODE 7590–01–M

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 May 20,

2000, through June 2, 2000. The last

biweekly notice was published on May

31, 2000.

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.

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37421

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 July 14, 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

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).

37422 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374,

LaSalle County Station, Units 1 and 2,

LaSalle County, Illinois

Date of amendment request: April 25,

2000.

Description of amendment request:

The proposed amendments would

revise Technical Specification (TS)

3/4.9.5, ‘‘Communications’’ to allow

movement of a control rod in a fueled

core cell in Operational Condition 5, to

be exempt from the communication

requirements of TS Section 3/4.9.5

when the control rod is moved with its

normal drive system.

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:

Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

TS Section 3/4.9.5 requires that direct

communications be maintained between the

control room and the refueling platform

personnel during Core Alterations in

Operational Condition 5. The requirement to

have direct communications maintained

between the control room and the refueling

platform personnel does not have an effect on

any accident previously evaluated or the

associated accident assumptions. Thus, the

proposed changes do not significantly

increase the probability of an accident

previously evaluated.

The proposed changes do not adversely

effect the integrity of the reactor coolant

system or secondary containment. As such,

the radiological consequences of previously

evaluated accidents are not changed.

Therefore, the proposed changes do not

increase the consequences of an accident

previously evaluated.

Does the change create the possibility of a

new or different kind of accident from any

accident previously evaluated?

The proposed changes do not affect the

assumed accident performance of any

structure, system, or component previously

evaluated. The proposed changes do not

introduce any new modes of system

operation or failure mechanisms.

Thus, these proposed changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Does the change involve a significant

reduction in a margin of safety?

LaSalle County Station, Units 1 and 2,

exercise control rods during Core Alterations

in Operational Condition 5. The required

plant conditions for this control rod

movement are specified in TS Section 3/

4.9.3, ‘‘Control Rod Position.’’ TS Section 3/

4.9.3 allows the movement of one control rod

at a time, in a fueled core cell, under control

of the reactor mode switch Refuel position

one-rod-out interlock. The exercising of

control rods under the control of the reactor

mode switch Refuel position one-rod-out

interlock is controlled by operators in the

control room and does not occur when fuel

is being moved in the reactor pressure vessel

(RPV).

The proposed changes do not affect the

margin of safety as the movement of a control

rod will continue to satisfy the requirements

of TS Section 3/4.9.3 and will not occur

when fuel is being moved in the RPV.

Thus, this 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

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374,

LaSalle County Station, Units 1 and 2,

LaSalle County, Illinois

Date of amendment request: April 28,

2000.

Description of amendment request:

The proposed amendments would

revise License Condition 2.C.(37) for

Unit 1 and License Condition 2.C.(21)

for Unit 2, to specify the types of fuel

movements that cannot be performed

during refueling unless all control rods

are fully inserted.

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:

Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes to LaSalle County

Station, Unit 1, License Condition 2.C.(37)

and Unit 2 License Condition 2.C.(21), will

require that control rods be fully inserted

during the loading and shuffling of fuel

assemblies during refueling in Operation

Condition 5. The requirement to have control

rods fully inserted during the loading or

shuffling of fuel assemblies, during a

refueling in Operational Condition 5, does

not have an effect on any accident previously

evaluated. The removal of fuel assemblies

from the RPV does not affect the initiators or

assumptions of a previously analyzed

accident, including inadvertent criticality.

Thus, the probability of the occurrence of an

accident previously evaluated is not

increased.

The proposed changes do not affect the

analyzed refueling accidents, the integrity of

the Reactor Coolant System or Secondary

Containment. Thus, the radiological

consequences of an accident previously

evaluated are not increased.

Therefore, the proposed changes do not

involve a significant increase in the

probability of occurrence or consequences of

an accident previously evaluated.

Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes to the Unit 1 and 2

License Conditions do not affect the assumed

accident performance of any structure,

system, or component previously evaluated.

The proposed changes do not introduce any

new modes of system operation or failure

mechanisms.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Does the change involve a significant

reduction in a margin of safety?

The shutdown margin required during a

refueling [outage] is specified in Technical

Specifications (TS) Section 3/4.1.1,

‘‘Shutdown Margin.’’ The required shutdown

margin ensures that the core will be

maintained sufficiently subcritical to

preclude inadvertent criticality in the

shutdown condition. The single failure

inadvertent criticality concerns, during a

refueling, are an unexpected withdrawal of a

control rod and the loading of a fuel

assembly into the wrong core cell location.

The analysis of these single failure

inadvertent criticality concerns, for a fully

loaded core, has determined that the most

limiting event is the unexpected withdrawal

of the highest worth control rod from a fueled

cell.

The proposed changes, to the Units 1 and

2 License Conditions, will prohibit the

loading and shuffling of any fuel assembly

within the RPV unless all control rods are

fully inserted during a refueling in

Operational Condition 5. The unloading of a

fuel assembly will be consistent with the fuel

assembly and control rod requirements of TS

Sections 3/4.9.10.1, ‘‘Single Control Rod

Removal,’’ and 3/4.9.10.2, ‘‘Multiple Control

Rod Removal.’’ These TS requirements

ensure that the proposed changes to the

license conditions will provide assurance

that the current analysis for an unexpected

withdrawal of the highest worth control rod

from a totally fueled core remains bounding

during a refueling outage.

Thus, 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

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37423

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374,

LaSalle County Station, Units 1 and 2,

LaSalle County, Illinois

Date of amendment request: May 1,

2000.

Description of amendment request:

The proposed amendments would

revise Technical Specification 3/4.8.1,

‘‘A. C. Sources—Operating,’’ to permit

functional testing of the emergency

diesel generators to be performed during

power operation.

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:

Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The function of the emergency diesel

generators (EDGs) is to supply emergency

power in the event of a loss of offsite power.

Operation of the EDGs is not a precursor to

any accident. Therefore, the proposed change

to permit the 24-hour functional test of the

EDGs to be performed during power

operation does not increase the probability of

an accident previously evaluated.

The EDG that is being tested will be

available to supply emergency loads within

the required time to mitigate an accident. In

addition, the remaining required EDGs will

be operable during the test. Furthermore,

with any one EDG inoperable the remaining

EDGs are capable of supporting the safe

shutdown of the plant. Therefore, the

consequences of an accident previously

evaluated are not significantly changed.

Therefore, the proposed changes will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes to the 24-hour

functional surveillance test will not affect the

operation of any safety system or alter its

response to any previously analyzed

accident. The EDG will automatically transfer

from the test mode of operation, if necessary,

to supply emergency loads in the required

time. This mode of operation is used for the

monthly surveillance of the EDGs. Therefore,

no new plant operating modes are

introduced.

In the event the EDG fails the functional

test, it will be declared inoperable and the

actions required for an inoperable EDG will

be performed. The remaining required EDGs

will be maintained operable and are capable

of feeding the loads necessary for safe

shutdown of the plant. This addresses the

concerns raised in the NRC Information

Notice 84–69, ‘‘Operation of Emergency

Diesel Generators,’’ regarding the operation

of EDG[s] connected in parallel with offsite

power. The Information Notice discusses

EDG configurations that have the potential to

lead to a complete loss of offsite and onsite

power to safety buses. In summary, the

proposed changes do not adversely affect the

performance or the ability of the EDGs to

perform their intended function.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Do the changes involve a significant

reduction in a margin of safety?

The proposed changes will not reduce

availability of the EDG being tested to

provide emergency power in the event of a

loss of offsite power. If a loss of offsite power

with a loss of coolant accident occurs during

the surveillance test, the emergency bus

would de-energize and shed load. The EDG

would then transfer from the test mode to the

emergency mode. It would then be available

to automatically supply emergency loads. In

addition, the remaining required EDGs would

be maintained operable during the test.

Furthermore, with any one EDG inoperable,

the remaining EDGs are capable of

supporting the safe shutdown of the plant.

The time required for the EDG being tested

to pick up emergency loads will not be

affected by performing the 24-hour functional

test during power operation.

The proposed changes do not affect the

assumptions or consequences of the analyzed

accidents. Therefore, the proposed changes

do not change any assumed safety margins.

Therefore, the proposed changes 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

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767

NRC Section Chief: Anthony J.

Mendiola

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of amendment request: April 13,

2000.

Description of amendment request:

The proposed amendment would revise

Technical Specification Surveillance

Requirement (SR) 3.3.1.1.10 for

Function 8 of Table 3.3.1.1–1 and SR

3.3.4.1.2.a. for reactor protection system

(RPS) and end of cycle (EOC)

recirculation pump trip instrumentation

to extend the frequency of these SRs

from 18 to 24 months.

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.

Actuation of the TTV [turbine throttle

valve] position switches is considered in the

Turbine Trip accident analysis in Chapter 15

of the WNP–2 Final Safety Analysis Report.

The valve position switches are assumed to

function normally at greater than 30% reactor

power level to initiate a reactor scram to

mitigate pressure increase and an RPT

[recirculation pump trip] to terminate jet

pump flow in the accident analysis. The

extension of the Channel Calibration

surveillance interval to 24 months does not

impact the normal function of the switches

that is assumed in the accident analysis.

There is no increase in probability or

consequences represented by the proposed

amendment.

Therefore, the extension of the surveillance

intervals does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Historical maintenance and surveillance

data indicate there is no effect on the

performance of the TTV position switches

resulting from an extension of the SR interval

from 18 to 24 months. To ensure reliability,

WNP–2 periodically replaces the TTV

position switches according to the

manufacturers’ recommendation. The

surveillance interval extension does not

involve a change in design or a change of

switch function. There is no increase in the

probability of failure expected from the

interval extension that could result in a

different kind of accident from any

previously evaluated.

Therefore, the operation of WNP–2 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. The proposed change does not involve

a significant reduction in a margin of safety.

Closure of the TTVs isolates the main

turbine as a heat sink producing reactor

pressure and neutron flux transients. Eight

TTV limit switches (two per valve) function

to actuate RPS and an EOC RPT to mitigate

these transients and terminate jet pump flow.

High pressure and flux transients also actuate

RPS resulting in negative reactivity insertion

should there be a failure of the TTV position

switches. Additionally, historical

maintenance and surveillance records

indicate that the TTV position switches will

operate within the necessary range and

accuracy with the extension of the SR

interval because no position adjustment has

been necessary during past TTV position

switch surveillance activities.

37424 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

Therefore, operation of WNP–2 in

accordance with the proposed amendment

will 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: Thomas C.

Poindexter, Esq., Winston & Strawn,

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

20005–3502

NRC Section Chief: Stephen Dembek

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: May 8,

2000.

Description of amendment request:

The proposed amendment would

change the River Bend Station, Unit 1

(River Bend or RBS), Technical

Specifications (TSs) to remove the Fuel

Building and the fuel building

ventilation system from the

requirements associated with the

Secondary Containment boundary

during operational Modes 1, 2, and 3.

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 to the Technical

Specifications involve removing the Fuel

Building and the fuel building ventilation

system from the requirements associated

with the Secondary Containment boundary.

The changes result in conservatively

assuming that all annulus bypass leakage

following a DBA [design basis accident]

LOCA [loss-of-coolant accident] are directed

to the environment for the duration of the

accident. Since the proposed changes only

affect functions that are required subsequent

to a LOCA or fuel handling accident (FHA),

the proposed changes have no [a]ffect on the

probability of an accident. The Fuel Building

portion of the Secondary Containment

boundary is not an active component that

could affect the proper operation of any other

essential safety feature or component.

Removal of the Fuel Building from the

Secondary Containment boundary does not

affect any other safety-related system,

component, or structure that would increase

the probability of an accident previously

evaluated. The proposed change only has an

impact on the dose consequences of the

design basis accident and does not have any

affect on the accident precursors or other

accident mitigating features.

A plant-specific radiological analysis has

been performed to assess the affects of the

proposed change in the annulus bypass

leakage release pathway in terms of Control

Room and off-site doses following a

postulated design basis LOCA. The

calculated doses for all offsite and onsite

evaluation points are within the 10 CFR

[Code of Federal Regulations] Part 100

criteria for offsite doses and within the

General Design Criterion 19 of 10 CFR Part

50 for the Control Room.

The calculated offsite DBA LOCA doses

due to the proposed changes result in an

increase of less than 3 percent due to

releasing all annulus bypass leakage directly

to the environment. The control room doses

exhibit the largest percentage increase in the

thyroid dose due to the increase in unfiltered

and untreated iodine released to the

environment, the release rate to the

environment, and the changes in the control

room atmospheric diffusion coefficient due

to dual air intakes. However, the change in

control room thyroid dose reduces the

margin to the regulatory limit by only 4

percent. The calculated doses for all offsite

and onsite evaluation points are not

significantly increased and remain within the

10 CFR Part 100 criteria for offsite doses and

within the General Design Criterion 19 of 10

CFR Part 50 for control room.

The proposed changes also include

relaxation of requirements for the fuel

building and fuel building ventilation system

except during the movement of ‘‘recently’’

irradiated fuel. The term ‘‘recently

irradiated’’ is defined as ‘‘fuel that has

occupied part of a critical reactor core within

the previous 11 days.’’ This change is

justified based on the irradiated fuel source

term decay period. River Bend currently

evaluates three FHA scenarios, one for the

fuel building and two for containment. The

FHA–FB [Fuel Building] scenario would be

impacted by the proposed changes since the

scenario assumed filtration for the duration

of the release. However, the proposed

changes are bounding in their entirety by the

FHA dose evaluation prepared in support of

Amendment 85, as revised to support

Amendment 110. The current analysis

assumes that a FHA occurs with the

containment personnel air locks (PAL) open,

thus, no credit is taken for primary

containment after an 11-day source term

decay period. The release rate assumed in

that analysis bounds the Fuel Building’s

normal ventilation rate by a factor of

approximately 3 and easily meets Regulatory

Guide 1.25 assumptions. All other data and

assumptions (other than decay time of

course) are identical for the two analyses and

thus, the Amendment 85 analysis is valid for

the Fuel Building.

It is therefore concluded that the proposed

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

(2) The operation of River Bend Station, in

accordance with the proposed amendment,

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes affect the TS

requirements for the fuel building and fuel

building ventilation system. These changes

have no impact on any other safety-related

system, component, or structure. The type of

accident and the accident precursors are not

affected by changing the annulus bypass

release path. The Fuel Building portion of the

Secondary Containment boundary is not an

active component that could affect the proper

operation of any other essential safety feature

or component. Also, the accident mitigating

features that are currently credited in the

response to the design basis accident are

unchanged by the proposed change.

Changing the release path for the annulus

bypass leakage does not create a new or

different kind of accident from the accidents

previously evaluated.

It is therefore concluded that the proposed

change does not create the possibility of a

new or different kind of accident from any

previously analyzed.

(3) The operation of River Bend Station, in

accordance with the proposed amendment,

does not involve a significant reduction in a

margin of safety.

The fuel building and the associated fuel

building ventilation filtration system are

currently credited as part of the secondary

containment function. The modified

secondary containment boundary (excluding

the fuel building) will still be capable of

performing its design function of limiting

offsite and control room dose to within

regulatory limits. The only accident

consequences that are impacted by the

proposed change in the secondary

containment (annulus) bypass leakage path

are the dose consequences of the design basis

LOCA. The previous dose analysis is changed

by assuming that all annulus bypass leakage

is directly to the environment instead of

being released into the Fuel Building where

the release would be treated by the Fuel

Building Ventilation System before release. A

plant-specific radiological analysis has been

performed to assess the affects of the

proposed change in the annulus bypass

leakage release pathway in terms of Control

Room and off-site doses following a

postulated design basis LOCA. The proposed

change required a revision to the existing

LOCA dose analysis since the annulus bypass

leakage release is assumed to be directly to

the environment due to removal of the Fuel

Building from the Secondary Containment

boundary. The calculated doses for all offsite

and onsite evaluation points are within the

10 CFR Part 100 criteria for offsite doses and

within the General Design Criterion 19 of 10

CFR Part 50 for the Control Room.

The proposed changes to the Technical

Specification requirements for the fuel

building and the fuel building ventilation

system when handling irradiated fuel in the

fuel building are bounded by currently

approved FHA analyses.

Therefore, there is no significant reduction

in the margin of safety associated with

postulated design basis events at RBS in

allowing the proposed change to the RBS

licensing basis.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37425

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: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC

20005.

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: July 29,

1999, as supplemented by letters dated

August 8, 1999, August 24, 1999,

January 27, 2000, March 29, 2000, May

22, 2000, and May 31, 2000.

Description of amendment request:

The proposed amendment request

provides additional information to

support a modification to Technical

Specification (TS) 3.8.1.1 and associated

Bases by extending the Emergency

Diesel Generator (EDG) allowed outage

time (AOT) from 72 hours to 10 days.

In the supplement letter dated May 22,

2000, an alternate source for the onsite

power system during the EDG

maintenance outage, by way of a

temporary EDG (TEDG) has been added.

The application dated July 29, 1999, did

not include the TEDG. This notice

supercedes the biweekly Federal

Register notice dated February 9, 2000,

(65 FR 6406) based on the original

application dated July 29, 1999.

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 this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response:

The EDGs are backup alternating current

power sources designed to power essential

safety systems in the event of a loss of offsite

power. As such, the EDGs are not accident

initiators in any accident previously

evaluated. Therefore, this change does not

involve a significant increase in the

probability of an accident previously

evaluated.

The proposed changes to the TS will

extend the allowed outage time (AOT) for a

single inoperable emergency diesel generator

(EDG) from the current limit of 72 hours to

10 days with the implementation of

compensatory measures. These compensatory

measures consist of a temporary emergency

diesel generator (TEDG) capable of supplying

auxiliary power to required safe shutdown

loads on the EDG train removed from service.

In the probabilistic risk assessment (PRA)

event of a loss of offsite power, the failure of

the operable EDG, and the failure of the

turbine-driven emergency feedwater pump to

start, the TEDG would be started and ready

for load within 25 minutes. In the PRA

assumptions to calculate the risk increase to

core damage, 50 minutes is available until

core uncovery. The AOT would be extended

for: (1) preplanned maintenance work (both

preventive and corrective) known to require

greater than 72 hours; and (2) unplanned

corrective maintenance work which may be

determined to take greater than 72 hours.

The plant defense-in-depth has been

preserved by the use of a TEDG to supply

required safe shutdown loads. The design

basis for the onsite power systems will

continue to conform to 10 CFR 50, Appendix

A, General Design Criterion 17.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Will the operation of the facility in

accordance with this proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response:

The EDGs are backup alternating current

power sources designed to power essential

safety systems in the event of a loss of offsite

power. The proposed changes to the TS will

extend the allowed outage time (AOT) for a

single inoperable emergency diesel generator

(EDG) from the current limit of 72 hours to

10 days with the implementation of

compensatory measures. These compensatory

measures consist of a temporary emergency

diesel generator (TEDG) capable of supplying

auxiliary power to required safe shutdown

loads on the EDG train removed from service.

In the PRA event of a loss of offsite power,

the failure of the operable EDG, and the

failure of the turbine-driven emergency

feedwater pump to start, the TEDG would be

started and ready for load within 25 minutes.

In the PRA assumptions to calculate the risk

increase to core damage, 50 minutes is

available until core uncovery. The AOT

would be extended for: (1) preplanned

maintenance work (both preventive and

corrective) known to require greater than 72

hours; and (2) unplanned corrective

maintenance work which may be determined

to take greater than 72 hours.

The proposed change does not alter the

design, configuration, and method of

operation of the plant for safety-related

equipment during the EDG AOT extension

period. The plant defense-in-depth has been

preserved by the use of a TEDG to supply

power to required safe shutdown loads.

The change does involve the modification

of non-safety permanent plant equipment.

The modification will involve preparing a

4.16kV [kilo-volt] non-safety bus breaker for

connection to the output of the TEDG. There

is no change being made to the parameters

within which the plant is operated, and the

setpoints at which the protective or

mitigative actions initiate. The design basis

on which the plant was licensed will not be

changed. In the PRA event of a loss of offsite

power, the failure of the operable EDG, and

the failure of the turbine-driven emergency

feedwater pump to start, the TEDG would be

started and ready for load within 25 minutes.

In the PRA assumptions to calculate the risk

increase to core damage, 50 minutes is

available until core uncovery.

Procedures will be developed to

implement onsite power system recovery

action in conjunction with the present

Emergency Operating Procedures (EOP) and

appropriate Off Normal Procedures in the

event it is necessary to use the alternate AC

power source. The developed procedures

support compensatory measures that provide

additional assurance that if a coincident Loss

of Offsite Power and failure of the operable

EDG (outside the design basis of the plant)

occurred during a preplanned maintenance

(both preventive and corrective) or

unplanned corrective maintenance extended

EDG AOT outage, appropriate guidance

would be available to safely shutdown the

plant. There are no alterations to the existing

plant procedure that will decrease assurance

that the plant will remain within analyzed

limits. As such, no new failure modes are

being introduced that would involve any

potential initiating events that would create

any new or different kind of accident. The

proposed change will only provide the plant

some flexibility in the AOT for

accomplishing preplanned maintenance

(both preventive and corrective) normally

performed during refueling outages and any

potential unplanned corrective maintenance

that may exceed the normal 72-hour AOT

during plant operation in Modes 1, 2, 3, and

4. The change does not alter assumptions

made in the safety analysis and licensing

basis.

Therefore, since there will be no

permanent hardware modifications to safetyrelated equipment nor alterations in the way

in which the plant or equipment is operated

during any design basis event, 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 this proposed change

involve a significant reduction in a margin of

safety?

Response:

The proposed change does not affect the

LCO’s [limiting conditions for operation] or

their Bases used in the deterministic analysis

to establish the margin of safety. The margin

of safety is established through equipment

design, operating parameters, and the

setpoints at which automatic actions are

initiated. There is no significant impact on

the margin of safety. PSA [probabilistic safety

assessment] methods were used to evaluate

the proposed change. The results of these

evaluations indicated the risk contribution

from this proposed AOT with compensatory

measures implemented during this extended

EDG AOT time period is small and within

the Regulatory Guide 1.177 risk-informed

acceptance guidelines.

Therefore, the change does not

significantly impact the margin of safety,

involve a permanent change in safety-related

plant design, or have any affect on the plant

protective barriers. Therefore, the proposed

change will not involve a significant

reduction in a margin of safety.

37426 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

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

29, 1999.

Description of amendment request:

Entergy Operations, Inc. (licensee) has

proposed to revise their Updated Final

Safety Analysis Report (UFSAR) to

discuss the probability threshold for

when physical protection of safetyrelated components from tornado

missiles is required for certain

components. The proposed changes

involve the use of Nuclear Regulatory

Commission (NRC) approved

probability risk methodology to assess

the need for additional tornado missile

protection and demonstrate that the

probability of damage due to tornado

missiles striking safety related

components is acceptably low.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes, i.e., revising the

current UFSAR descriptions addressing

tornado missile barrier protection at

Waterford Steam Electric Station, Unit 3

(Waterford 3) have been evaluated against

these three criteria, and it has been

determined that the changes do not involve

a significant hazard because:

(1) The proposed activity does not involve

a significant increase in the probability or

consequences of any accident previously

evaluated.

The associated UFSAR changes reflect use

of the Electric Power Research Institute

(EPRI) Topical Report, ‘‘Tornado Missile Risk

Evaluation Methodology, (EPRI NP–2005),’’

Volumes 1 and 2. This methodology has been

reviewed, accepted and documented in a

NRC Safety Evaluation dated October 26,

1983. The NRC concluded that: ‘‘the EPRI

methodology can be utilized when assessing

the need for positive tornado missile

protection for specific safety-related plant

features in accordance with the criteria of

SRP [Standard Review Plan] Section 3.5.1.4.’’

The EPRI methodology has been previously

applied by other licensees to resolve tornado

missile protection issues.

The results of the tornado missile hazards

analysis are such that the calculated total

tornado missile hazard probability for safetyrelated SSC’s [systems, structures and

components] is approximately 6.0 × 10 ¥7

per year. This is lower than the value

determined to be acceptable, i.e., 1 × 10¥6

per year by the NRC Staff.

With respect to the probability of

occurrence or the consequences of an

accident previously analyzed in the UFSAR,

the probability of a tornado reaching

Waterford 3 causing damage to plant systems,

structures and components is a design basis

event considered in the UFSAR. The changes

being proposed herein do not reduce the

probability that a tornado will reach the

plant. However, it was determined that there

are a limited number of safety-related

components that theoretically could be

struck. The probability of tornado-generated

missile strikes on these components were

analyzed using the NRC Staff approved

probability methods described above. On this

basis, the proposed change is not considered

to constitute a significant increase in the

probability of occurrence or the

consequences of an accident, due to the low

probability of a tornado missile striking these

components.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of previously

evaluated accidents.

(2) The proposed activity does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes involve evaluation

of whether any physical protection of safetyrelated equipment from tornado missiles is

required relative to the probability of such

damage without physical protection. A

tornado at Waterford 3 is a design basis event

considered in the UFSAR. This change

involves recognition of the acceptability of

performing tornado missile probability

calculations in accordance with established

regulatory guidance.

Therefore, the change would not contribute

to the possibility of, or be the initiator for any

new or different kind of accident, or to occur

coincident with any of the design basis

accidents in the UFSAR. The low probability

threshold established for tornado missile

damage to system components is consistent

with these assumptions.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident.

(3) The proposed activity does not involve

a significant reduction on a margin of safety.

The request does not involve a significant

reduction in a margin of safety. The existing

licensing basis for Waterford 3 with respect

to the design basis event of a tornado

reaching the plant, generating missiles and

directing them toward safety-related systems

and components is to provide positive

missile barriers for all safety-related systems

and components. With the change, it will be

recognized that there is an extremely low

probability, below an established acceptance

limit, that a limited subset of the ‘‘important’’

systems and components could be struck.

The change from ‘‘protecting all safetyrelated systems and components’’ to ‘‘an

extremely low probability of occurrence of

tornado generated missile strikes on portions

of important systems and components’ is not

considered to constitute a significant

decrease in the margin of safety due to that

extremely low probability.

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: N. S. Reynolds,

Esquire, Winston & Strawn 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–334,

Beaver Valley Power Station, Unit No. 1,

Shippingport, Pennsylvania

Date of amendment request: February

21, 2000.

Description of amendment request:

The proposed amendment would revise

the Unit 1 Updated Final Safety

Analysis Report (UFSAR) descriptions

for bolting material used on some

Reactor Coolant System (RCS)

components.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The use of carbon steel fasteners in a

borated system introduces a new failure

mechanism for the fasteners, that of boric

acid wastage. The materials currently

specified in the [Beaver Valley Power

Station] BVPS Unit 1 UFSAR are not

susceptible to boric acid wastage. The

probability of failure for all systems may be

increased due to the additional failure mode

introduced by change from corrosion

resistant material to carbon steel for RCS and

reactor coolant pressure boundary fasteners.

The design requirements of the [American

National Standards Institute] ANSI and

[American Society of Mechanical Engineers]

ASME Codes are conservative in nature, in

that, the stress allowable for fastener

materials is less than half the yield strength

of the material, thus creating a margin in the

design of two or greater on structural

strength. Therefore, the failure or damage of

one or more non-adjacent fasteners can

normally be accommodated. Additionally,

the material properties (Yield and Tensile

strength) of the installed (SA540 Grade B

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37427

Class 23 or 24) carbon steel fasteners are

higher than that of the material identified in

the UFSAR (SA453 Grade 660). It should also

be noted that the use of either the carbon

steel fasteners (those installed) or the

stainless steel fasteners (those identified in

the UFSAR) is acceptable by the design

Codes (ANSI and ASME), the selection of the

material for the fasteners is at the discretion

of the designer and is not specified by Code

requirements. When compared to carbon

steel fasteners, the corrosion resistance of

Grade 660 material is pertinent only if

leakage is actively occurring.

The boric acid wastage concern is

mitigated by the Boric Acid Corrosion

Program which has systematic measures to

ensure that boric acid corrosion will not lead

to degradation of the reactor coolant pressure

boundary. This Boric Acid Corrosion

Program with its inspections provides

adequate assurances that abnormal leakage

will be identified and corrective actions

taken prior to significant boric acid corrosion

degradation of carbon steel pressure

boundary components.

The NRC, in Generic Letter (GL) 88–05,

recognized that boric acid solution leaking

from the reactor coolant system can cause

significant corrosion damage to carbon steel

materials. In the GL, the NRC requested that

licensees provide assurance that a boric acid

monitoring program has been implemented.

This program was to consist of systematic

measures to ensure that boric acid corrosion

does not lead to degradation of the assurance

that the reactor coolant pressure boundary

will have an extremely low probability of

abnormal leakage or rupture. The Beaver

Valley Power Station response to the GL

provided assurance that a program was in

place and committed to enhancements to the

existing program. An NRC follow-up review

was conducted and the Beaver Valley Power

Station program was found to be acceptable

and fulfilling the requirements of GL 88–05

(Reference: NRC Inspection Report Nos. 50–

334/88–23 and 50–334/88–25)

Therefore, the proposed changes to BVPS

Unit 1 UFSAR Tables 1.8–1 and 1.8–2 do not

significantly increase the probability or

consequences of any accident previously

evaluated in the BVPS Unit 1 UFSAR.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

During an evaluation of the fastener

material to be used for the replacement of a

degraded fastener, it was discovered that the

BVPS Unit 1 UFSAR Tables 1.8–1 and 1.8–

2 identified that corrosion resistant materials,

SA453 Grade 660, were identified as being

installed. The use of carbon steel fasteners in

lieu of the SA453 Grade 660 fasteners

identified in the UFSAR introduces the

potential failure mechanism of boric acid

corrosion. The corrosion damage that has

occurred on MOV–RC–591 and MOV–CH–

310 bolting demonstrates that corrosion

damage from unchecked borated water

leakage is damaging to carbon steel fasteners.

Additionally, it should be noted that both of

these degraded conditions were identified

and repaired prior to an operational or

structural concern through the application of

the Boric Acid Corrosion Program.

In the design condition (non-corroded), the

change to carbon steel fasteners would not

affect the design basis accidents described in

the UFSAR. The boric acid wastage concern

is mitigated by the Boric Acid Corrosion

Program which has systematic measures to

ensure that boric acid corrosion will not lead

to degradation of the reactor coolant pressure

boundary.

In addition to the Boric Acid Corrosion

Program, the body to bonnet configuration for

the fasteners identified in Table 1.8–1 and

1.8–2 result in multiple fasteners for each

joint. To meet the requirements of the design

Codes (ANSI or ASME) for valves, the

number of fasteners installed is in excess of

the number of fasteners required to perform

the structural function of maintaining the

pressure boundary. Additionally, it is highly

unlikely that all the installed fasteners would

corrode in such a manner that catastrophic

failure of the body to bonnet joint would

result. Therefore, the multiple installed

fasteners result in an installed backup to the

minimum required number of fasteners

necessary to maintain pressure boundary

integrity.

Thus, the assumptions and consequences

of the loss of pressure boundary integrity

type of accident would be unchanged and

would not introduce a new or different kind

of accident as currently evaluated in the

BVPS Unit 1 UFSAR based on the Boric Acid

Corrosion Program preventing any

unacceptable boric acid wastage in

accordance with GL 88–05.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed change in the Unit 1 UFSAR

removing criteria requiring stainless steel

fasteners for RCS and reactor coolant

pressure boundary components would not

involve a significant reduction in the margin

of safety since current Technical

Specification requirements remain

unchanged and current plant programs (i.e.,

Boric Acid Corrosion Program inspections)

provide adequate assurance from the

likelihood of corroded fasteners causing an

operational issue. NRC reviewed the Beaver

Valley Power Station Boric Acid Corrosion

Program and found the program to be

acceptable to fulfill the requirements of GL

88–05 (Reference: NRC Inspection Report

Nos. 50–334/88–23 and 50–334/88–25).

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: Mary O’Reilly,

FirstEnergy Nuclear Operating

Company, FirstEnergy Corporation, 76

South Main Street, Akron, OH 44308.

NRC Acting Section Chief: Marsha

Gamberoni.

FirstEnergy Nuclear Operating

Company, et al., Docket Nos. 50–334

and 50–412, Beaver Valley Power

Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: May 1,

2000.

Description of amendment request:

The proposed amendment would revise

the Unit 1 and 2 Technical Specification

(TS) 3/4.6.4.2 Surveillance Requirement

(SR). The proposed change would allow

performance of the hydrogen

recombiner functional test at

containment pressures greater than the

currently specified 13 psia. This would

be accomplished by measuring the flow

under normal or current test conditions

(e.g., atmospheric pressure) and

calculating the expected system

performance under design basis

operating conditions. The surveillance

would be revised to verify that the

recombiner flow, when corrected to the

post accident design conditions, is

greater than or equal to the required

flow. The corresponding design basis

temperature for post accident

recombiner operation would be

included in the SR because it is required

to correct the test flow to the design

basis operating conditions. In order to

support the calculations necessary to

confirm the recombiner blower

performance, the proposed change

includes the addition of an equation and

associated discussion to the bases. The

equation will correct the measured test

flow to a corresponding flow at the

design basis operating pressure and

temperature. In addition to the technical

change described above, SR 4.6.4.2.b.3

would be modified by separating the

criteria for the system blower

performance and heater operation into

separate parts of the same surveillance

to improve the presentation of the

requirements. Format and editorial

changes are included as necessary to

facilitate the revision of the TS text to

conform to the current TS page format,

and addition of text to the bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change does not result in

any hardware changes to the hydrogen

recombiners. Additionally, the hydrogen

recombiners are not assumed to be accident

initiators of any analyzed event. The

proposed change revises the method for

performing the hydrogen recombiner

37428 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

functional test specified in Technical

Specification (TS) Surveillance Requirement

(SR) 4.6.4.2.b.3. The proposed change to SR

4.6.4.2.b.3 does not reduce the effectiveness

of the requirement and continues to verify

the capability of the hydrogen recombiners to

perform their design basis function

consistent with the assumptions of the

applicable safety analysis. Therefore, the

consequences or probability of accidents

previously evaluated in the UFSAR remain

unchanged.

The addition of supporting TS bases text

and the format and editorial changes made to

the TS have no impact on plant operation or

safety.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Does the change create the possibility

of a new or different kind of accident from

any accident previously evaluated?

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change does

not affect any accidents previously evaluated

in the UFSAR and continues to provide

assurance that the hydrogen recombiners

remain capable of performing their design

function. The proposed change does not

introduce any new failure modes or affect the

probability of a malfunction.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Does the change involve a significant

reduction in a margin of safety?

The margin of safety depends on the

maintenance of specific operating parameters

and systems within design requirements and

safety analysis assumptions.

The proposed change does not involve

revisions to any safety limits or safety system

settings that would adversely impact plant

safety. In addition, the proposed change does

not affect the ability of the hydrogen

recombiners to perform their design function.

The proposed change revises the method

for performing the hydrogen recombiner

functional test specified in SR 4.6.4.2.b.3.

However, the proposed change to SR

4.6.4.2.b.3 does not reduce the effectiveness

of the requirement and continues to verify

the capability of the hydrogen recombiners to

perform their design basis function

consistent with the assumptions of the

applicable safety analysis.

The addition of supporting TS bases text

and the format and editorial changes made to

the TS have no impact on plant operation or

safety.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mary O’Reilly,

FirstEnergy Nuclear Operating

Company, FirstEnergy Corporation, 76

South Main Street, Akron, OH 44308.

NRC Acting Section Chief: Marsha

Gamberoni.

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: April 27,

2000.

Description of amendment request:

The proposed amendment would

change the James A. FitzPatrick Nuclear

Power Plant Technical Specifications by

changes to the Trip Level Settings for

the Residual Heat Removal (RHR) and

Core Spray (CS) Pump Start Timers as

well as the Automatic Depressurization

System (ADS) Auto-Blowdown Timer.

The amendment would also extend the

Logic System Functional Test

surveillance test intervals for the RHR,

CS and ADS systems from 6 months to

24 months.

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,

since it would not:

Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This proposed change revises the Trip

Level Settings for the RHR and CS pump

interlock start timers as well as the ADS autoblowdown timers. This proposed change also

extends the surveillance interval for these

timers from 6-months to 24-months.

This proposed change impacts the control

of systems designed to mitigate the

consequences of a Loss of Coolant Accident

(LOCA). These changes do not impact any of

the Reactor Coolant System parameter

variations listed as potential causes of threats

to the fuel and Reactor Coolant Pressure

Boundary listed in section 14.4.2 of the

FitzPatrick UFSAR [Updated Final Safety

Analysis Report] (Reference 8) [see

application dated April 27, 2000]. Therefore,

this proposed change does not increase the

probability of an accident previously

evaluated.

The changes to the control of systems

designed to mitigate the consequences of

postulated LOCA events are consistent with

the relevant assumptions made in the

FitzPatrick LOCA analysis (Reference 5) [see

application dated April 27, 2000]. Therefore,

the results of that analysis are not changed.

Therefore, this proposed change does not

increase the consequence of an accident

previously evaluated. Create the possibility

of a new or different kind of accident from

any accident previously evaluated.

This proposed change impacts the control

of systems designed to mitigate the

consequences of a Loss of Coolant Accident

(LOCA). These changes do not impact any of

the Reactor Coolant System parameter

variations listed as potential causes of threats

to the fuel and Reactor Coolant Pressure

Boundary listed in section 14.4.2 of the

FitzPatrick UFSAR (Reference 8) [see

application dated April 27, 2000]. Therefore,

this proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Involve a significant reduction in a margin

of safety.

The changes to the control of systems

designed to mitigate the consequences of

postulated LOCA events are consistent with

the relevant assumptions made in the

FitzPatrick LOCA analysis (Reference 5) [see

application dated April 27, 2000]. Therefore

the results of that analysis are not changed.

Therefore, this proposed change does not

reduce 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: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Section Chief: Marsha

Gamberoni, Acting.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Units Nos.

1 and 2, Salem County, New Jersey

Date of amendment request: February

7, 2000.

Description of amendment request:

The proposed amendments would

revise Technical Specification 4.7.1.2.b

to make the surveillance requirements

for Auxiliary Feedwater Pump testing

consistent with that of NUREG–1431,

‘‘Standard Technical Specifications,

Westinghouse Plants.’’ The Bases

associated with this Technical

Specification would also be revised to

address the proposed changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to the Technical

Specification surveillance requirements for

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37429

the auxiliary feedwater pumps surveillance

testing are consistent with the latest auxiliary

feedwater flow hydraulic model and accident

analyses. The revised minimum acceptance

criteria will ensure that pump degradation,

which could adversely impact the accident

analyses, will be detected. The pumps will

continue to operate in the same manner as

assumed in the analyses to mitigate the

design basis accidents.

Therefore, there will be no significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes to the Technical

Specification surveillance requirements for

the auxiliary feedwater pumps surveillance

testing are consistent with the latest auxiliary

feedwater flow hydraulic model and accident

analyses. The proposed changes to the

Technical Specification surveillance

requirements and associated Bases will not

affect the way the pumps are operated during

normal plant operations or how the pumps

will operate after an accident.

Therefore, the proposed changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes to the Technical

Specification surveillance requirements for

the auxiliary feedwater pumps surveillance

testing are consistent with the latest auxiliary

feedwater flow hydraulic model and accident

analyses. The proposed changes to the

Technical Specification surveillance

requirements eliminate a potential nonconservative acceptance value and establish

appropriate restrictions to ensure pump

operability. The proposed change to the

Technical Specifications Bases better

describes the design function of the auxiliary

feedwater system.

Therefore, there will be no significant

reduction in the margin of safety as defined

in the Bases for the Technical Specifications

affected by these proposed changes.

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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Section Chief: James W. Clifford.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: May 25,

2000 (ULNRC–04257).

Description of amendment request:

The licensee proposed to eliminate the

technical specifications (TSs) on the

boron dilution mitigation system to

avoid a spurious swapover event, such

as occurred during the shutdown for

Refueling Outage 9, about 2 years ago.

This amendment would delete the

limiting condition for operation, the

actions, and the surveillance

requirements for TS 3.3.9, ‘‘Boron

Dilution Mitigation System (BDMS),’’ in

the instrumentation section of the TSs

for Callaway. In addition, the title of TS

3.3.9 would be removed from the Table

of Contents, the Bases for the TSs would

be revised, and a section on the boron

dilution analysis would be added to

Chapter 16 of the Callaway Final Safety

Analysis Report (FSAR).

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.

Overall protection system performance will

remain within the bounds of the previously

performed accident analyses since the

associated hardware changes described in

Section X of Appendix A [to the application

dated May 25, 2000] do not affect any

protection systems. The RTS [reactor trip

system] and ESFAS [engineered safety

features actuation system] instrumentation

will be unaffected. These protection systems

will continue to function in a manner

consistent with the plant design basis. The

installation of an alarm on the [reactor

coolant] letdown divert valve, addition of

two redundant high VCT [volume control

tank] water level alarms, and elimination of

the automatic BDMS valve swap-over

function will be performed in such a manner

that all design, material, and construction

standards that were applicable prior to the

change are maintained.

The proposed change will modify the

system interface between CVCS [chemical

and volume control system] and the boron

recycle system such that the RCS [reactor

coolant system] and CVCS form a closed

system consistent with the reanalysis

assumptions. The letdown divert valve will

be placed in the manual ‘‘VCT’’ mode [(1)]

prior to entry into MODE 3 from MODE 2

during a plant shutdown and [(2)] prior to

entry into MODE 5 from MODE 6 during a

plant startup such that letdown flow is

directed to the VCT, rather than to the

recycle holdup tanks, except under

administrative controls for planned

evolutions which require a high degree of

operator involvement and awareness. These

administrative controls will include

verification of the boron concentration of the

makeup [to the reactor coolant] prior to

repositioning the divert valve and restoration

requirements to return the valve to the

manual ‘‘VCT’’ mode upon evolution

completion.

The proposed change will not affect the

probability of any event initiators. The above

modifications are unrelated to the initiating

event for this analysis, a failure in the reactor

makeup control system. The change will

revise the method of detecting the event and

rely on operator action for event termination.

There will be no degradation in the

performance of or an increase in the number

of challenges imposed on safety-related

equipment assumed to function during an

accident situation. There will be no change

to normal plant operating parameters or

accident mitigation performance.

Since manual operator actions are being

substituted for automatic actions, this

amendment application was reviewed against

the guidance provided in NRC Information

Notice 97–78, ‘‘Crediting of Operator Actions

in Place of Automatic Actions and

Modifications of Operator Actions, Including

Response Times.’’ Appendix A [to the

application] demonstrates that sufficient time

is available for operator action to terminate

the inadvertent boron dilution event prior to

criticality. Additionally, as discussed in

NSAC–183, ‘‘Risk of PWR Reactivity

Accidents during Shutdown and Refueling,’’

gradual inadvertent boron dilution events are

not expected to cause core damage, even if

they are unmitigated, due to their selflimiting nature.

The proposed change will achieve the

same objective as the BDMS, i.e., the

prevention of an inadvertent criticality as a

result of an unintended boron dilution. The

proposed change will not alter any

assumptions or change any mitigation actions

in the radiological consequence evaluations

in the FSAR. Appendix A [to the application]

demonstrates that sufficient time is available

for operator action to terminate the

inadvertent boron dilution event prior to

criticality. With the reactor subcritical, there

will be no increase in radiological

consequences.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There are no changes in the method by

which any safety-related plant system

performs its safety function. The changes

described in Section X of Appendix A [to the

application] have no impact on any analyzed

event other than inadvertent boron dilution.

The physical modifications to eliminate the

automatic BDMS valve swap-over function

and add redundant high VCT water level

alarms and a position alarm on the letdown

divert valve will be implemented in

accordance with existing plant design

criteria. The BDMS itself has no impact on

any other analyzed event. The portion of the

change deleting the BDMS from the

Technical Specifications, and eliminating the

automatic valve swap-over function, has no

other impact safety. The BDMS flux

multiplication alarm will be retained as a

plant design feature to provide the plant

operators a diverse method for identifying a

potential dilution event. Since the passive

37430 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

alarms to be added only provide information

and do not initiate control or protection

system actions, the alarms will not adversely

impact other events. The position of the

letdown divert valve only affects the path for

letdown flow. The flow path selected for

letdown does not affect any other accident

analyses. Thus, the operational change to

make the manual ‘‘VCT’’ mode the normal

operating condition in MODES 3 through 5

has no safety impact. Procedural changes will

heighten the operator awareness of potential

dilution events and provide alarm response

actions to mitigate potential dilution events.

As such, these changes will enhance the

response to inadvertent boron dilution

events, but have no other safety impact. The

FSAR Chapter 16 requirements for reactor

coolant loop operation and high VCT water

level alarm operability will also enhance the

plant operators’ capability to respond to an

inadvertent boron dilution event. If the

Chapter 16 requirements are not met,

isolating the dilution source valves in

MODES 3, 4, and 5 has no impact on any

other accident analyses since none of the

other accident analyses take credit for, or are

initiated by, the flow path through these

valves.

This change will affect the normal method

of plant operation while in MODES 3 through

5 with regard to the control of letdown flow.

In these MODES, letdown processing via the

recycle holdup tanks will be allowed only

under administrative controls for planned

evolutions which require a high degree of

operator involvement and awareness. The

annunication of the letdown divert valve not

being in the ‘‘VCT’’ position will further

highlight system conditions to the operating

staff. No other performance requirements will

be affected.

In order to automatically close the VCT

isolation valves, the RWST [refueling water

storage tank] isolation valves must be fully

open. This valve interlock feature is designed

to ensure a flow path is maintained to the

CCPs [component cooling pumps] during

swap-over. Since the VCT isolation valves

can be manually closed prior to opening the

RWST isolation valves, the possibility exists

for the operator to inadvertently isolate flow

to the CCPs while attempting to isolate the

dilution source. However, plant operating

procedures provide the operators with

sufficient guidance for performing a manual

valve swap-over and the reanalysis

demonstrates that sufficient time is available

to perform the required manual actions,

consistent with SRP [NRC NUREG–0800

Standard Review Plan] acceptance criteria.

Therefore, the proposed change does 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 change uses acceptance

criteria consistent with the [NRC] Standard

Review Plan, as discussed in Appendix A [to

the application]. The margin of safety

required of the BDMS is maintained, i.e.,

inadvertent boron dilution events will be

terminated by timely operator actions prior to

a total loss of all shutdown margin. There

will be no effect on the manner in which

safety limits or limiting safety system settings

are determined nor will there be any effect

on those plant systems necessary to assure

the accomplishment of protective functions.

There will be no impact on the overpower

limit, DNBR [departure from nucleate boiling

ratio] limits, FQ, FdeltaH, LOCA PCT [loss-ofcoolant accident peak cladding temperature],

peak local power density, or any other

margin of safety. The radiological dose

consequences acceptance criteria listed in the

Standard review Plan will continue to be

met.

Therefore, the proposed change does not

involve a significant reduction in any margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied.Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: John O’Neill,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Section Chief: Stephen Dembek.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: May 22,

2000.

Description of amendment request:

The proposed amendment would

remove the technical specification

surveillance requirement for visual

inspection of suppression chamber

coating integrity once each refueling

outage.

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 change conforms the TS to

current regulations, credits actions taken

under GL 98–04 to address coating

delamination concerns, and eliminates

redundant surveillance criteria. Since reactor

operation under the revised Specification is

unchanged, no design or analytical

acceptance criteria will be exceeded. As

such, this change does not impact initiators

of analyzed events or assumed mitigation of

accident or transient events. The structural

and functional integrity of plant systems is

unaffected. Thus, there is no significant

increase in the probability or consequences

of accidents 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 change does not affect any

parameters or conditions that could

contribute to the initiation of any accident.

No new accident modes are created. No

safety-related equipment or safety functions

are altered as a result of these changes.

Because it does not involve any change to the

plant or the manner in which it is operated,

the proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

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.

The proposed change does not affect

design margins or assumptions used in

accident analyses and has no effect on any

initial condition. The capability of safety

systems to function and limiting safety

system settings are similarly unaffected as a

result of this change. Thus, the margins of

safety required for safety analyses are

maintained.

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.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: May 23,

2000.

Description of amendment request:

This proposed change relocates those

portions of Technical Specifications

(TSs) related to reactor coolant

conductivity and chloride requirements

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:

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 change is administrative in

nature and does not involve the modification

of any plant equipment or affect basic plant

operation. Conductivity and chloride limits

are not assumed to be an initiator of any

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37431

analyzed event, nor are these limits assumed

in the mitigation of consequences of

accidents.

Therefore, the proposed change does 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 change does not involve any

physical alteration of plant equipment and

does not change the method by which any

safety-related system performs its function.

As such, no new or different types of

equipment will be installed, and the basic

operation of installed equipment is

unchanged. The methods governing plant

operation and testing remain consistent with

current safety analysis assumptions.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

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.

The proposed change represents the

relocation of current Technical Specification

requirements to the Technical Requirements

Manual, based on regulatory guidance and

previously approved changes for other

stations. The proposed change is

administrative in nature, does not negate any

existing requirement, and does not adversely

affect existing plant safety margins or the

reliability of the equipment assumed to

operate in the safety analysis. As such, there

are no changes being made to safety analysis

assumptions, safety limits or safety system

settings that would adversely affect plant

safety as a result of the proposed change.

Margins of safety are unaffected by

requirements that are retained, but relocated

from the Technical Specifications to the

Technical Requirements Manual.

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. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: May 23,

2000.

Description of amendment request:

The proposed amendment would revise

the technical specification surveillance

requirements for local power range

monitor calibration.

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 revised surveillance requirement

continues to ensure that the local power

range monitor (LPRM) signal is adequately

calibrated. This change will not alter the

basic operation of process variables,

structures, systems, or components as

described in the safety analyses, and no new

equipment is introduced by the change in

LPRM surveillance interval. Therefore, the

probability of accidents previously evaluated

is unchanged.

The consequences of an accident can be

affected by the thermal limits existing at the

time of the postulated accident, but LPRM

chamber exposure has no significant effect on

the calculated thermal limits because LPRM

accuracy does not significantly deviate with

exposure. For the extended calibration

interval, the total nodal power uncertainty

remains less than the uncertainty assumed in

the thermal analysis basis safety limit,

maintaining the accuracy of the thermal limit

calculation. Therefore, the thermal limit

calculation is not significantly affected by

LPRM calibration frequency, and the

consequences of an accident previously

evaluated are unchanged.

These changes do not affect the initiation

of any event, nor do they negatively impact

the mitigation of any event. 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 change will not physically

alter the plant or its mode of operation. As

such, no new or different types of equipment

will be installed, and the basic operation of

installed equipment is unchanged. The

methods governing plant operation and

testing are consistent with current safety

analysis assumptions. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

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

fundamental operation, and there are no

changes being made to safety limits or safety

system settings that would adversely affect

plant safety as a result of the proposed

change. The margin of safety can be affected

by the thermal limits existing prior to an

accident; however, uncertainties associated

with LPRM chamber exposure have no

significant effect on the calculated thermal

limits. The thermal limit calculation is not

significantly affected because LPRM

sensitivity with exposure is well defined.

LPRM accuracy remains within the total

nodal power uncertainty assumed in the

thermal analysis basis, thus maintaining

thermal limits and the safety margin.

Since the proposed changes do not affect

safety analysis assumptions or initial

conditions, the margins of safety in the safety

analyses are maintained. 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.

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.

AmerGen Energy Company, LLC,

Docket No. 50–289, Three Mile Island

Nuclear Station, Unit 1, Dauphin

County, Pennsylvania

Date of amendment request: May 4,

2000, as supplemented May 9, 2000.

Description of amendment request:

The proposed amendment would

remove the individual control building

isolation and recirculation damper

numbers from Technical Specification

4.12.1.3 and instead specify ‘‘required’’

37432 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

dampers. The requirement to test these

dampers remains the same. The Bases

have been modified to indicate that the

damper numbers for control building

isolation and recirculation are contained

in the Updated Final Safety Analysis

Report.

Date of publication of individual

notice in Federal Register: May 22, 2000

(65 FR 32132).

Expiration date of individual notice:

June 21, 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, et

al., Docket No. 50–325, Brunswick

Steam Electric Plant, Unit 1, Brunswick

County, North Carolina

Date of application for amendment:

April 14, 2000, as supplemented April

20, 2000.

Brief description of amendment: The

amendment changed Technical

Specification Surveillance Requirement

3.1.3.3 to allow partial insertion of

control rod 26–47 instead of insertion of

one complete notch. This revised

acceptance criterion is limited to the

current Unit No. 1 operating cycle, after

which the original one-notch

requirement will be re-established.

Date of issuance: May 23, 2000.

Effective date: May 23, 2000.

Amendment No.: 210.

Facility Operating License No. DPR–

71: Amendment changes the Technical

Specifications.

Date of initial notice in Federal

Register: April 21, 2000 (65 FR 21481).

The April 20, 2000, supplemental letter

contained clarifying information only,

and did not change the initial no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 23, 2000.

No significant hazards consideration

comments received: No.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of application for amendment:

July 29, 1999, as supplemented by letter

dated January 31, 2000.

Brief description of amendment: The

amendment revised Technical

Specification (TS) 3.4.9 applicability

from Mode 3 with steam dome pressure

less than residual heat removal cut in

permissive to Mode 3 with steam dome

pressure less than 48 psig. Notes

associated with TS Surveillance

Requirements 3.4.9.1 and 3.5.1.2 are

changed to reflect the new 48 psig limit.

Date of issuance: May 23, 2000.

Effective date: May 23, 2000, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 164.

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR

46430).

The January 31, 2000, 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 amendment is contained in a

Safety Evaluation dated May 23, 2000.

No significant hazards consideration

comments received: No.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of application for amendment:

July 29, 1999, as supplemented by

letters dated August 30, 1999, and

February 28, 2000.

Brief description of amendment: The

amendment deletes item 3.(b) of

Attachment 2 to License Condition

2.C.(16), that required installation of a

neutron flux monitoring system, in the

form of excore wide range monitors, in

conformance with Regulatory Guide

1.97, ‘‘Instrumentation for Light-WaterCooled Nuclear Power Plants to Assess

Plant and Environs Conditions During

and Following an Accident.’’

Date of issuance: May 18, 2000.

Effective date: May 18, 2000, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 162.

Facility Operating License No. NPF–

21: The amendment revised the

Operating License.

Date of initial notice in Federal

Register: October 20, 1999 (64 FR

56530).

The February 28, 2000, supplemental

letter provided additional clarifying

information but 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 amendment is contained in a

Safety Evaluation dated May 18, 2000.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of application for amendment:

November 18, 1999, as supplemented by

a letter dated February 7, 2000.

Brief description of amendment: The

amendment revised Subsection 4.3.1.2.b

of Technical Specification 4.3, ‘‘Fuel

Storage.’’ The change revised the

wording which described the spacing of

the fuel in the new fuel racks.

Date of issuance: May 23, 2000.

Effective date: May 23, 2000, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 163.

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37433

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73088)

The February 7, 2000, 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 amendment is contained in a

Safety Evaluation dated May, 23, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No.

50–368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

November 29, 1999.

Brief description of amendment: The

amendment relocated the requirements

associated with the high steam generator

level trip functions of the Reactor

Protection System from the Technical

Specifications to the Technical

Requirements Manual.

Date of issuance: May 18, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment No.: 216.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 9, 2000 (65 FR 6404).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 18, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: August 4,

1999, as supplemented by letter dated

May 18, 2000.

Brief description of amendment: The

proposed change modifies the Technical

Specifications (TS) to extend allowed

outage time (AOT) to seven days for one

inoperable low pressure safety injection

(LPSI) train. Additionally, an AOT of 72

hours is imposed for other conditions

where the equivalent of 100 percent

emergency core cooling system (ECCS)

subsystem flow is available. If 100

percent ECCS flow is unavailable due to

two inoperable LPSI trains, an ACTION

has been added to restore at least one

LPSI train to OPERABLE status within

one hour or place the plant in HOT

STANDBY within six hours, and to exit

the MODE of applicability within the

following six hours. In the event the

equivalent of 100 percent ECCS

subsystem flow is not available due to

other conditions, TS 3.0.3 is entered.

The Limiting Condition for Operation

terminology is being changed for

consistency with the ECCS

requirements. Additionally, the

associated TS Bases are being changed.

Date of issuance: May 25, 2000.

Effective date: As of the date of

issuance and shall be implemented 60

days from the date of issuance.

Amendment No.: 164.

Facility Operating License No. NPF–

38: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4278).

The May 18, 2000, supplement did

not expand the scope of the application

as noticed or change the proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 25, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, Docket No. 50–440, Perry

Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of application for amendment:

September 9, 1999, as supplemented by

submittals dated March 1, March 13,

and May 11, 2000.

Brief description of amendment: This

amendment increases the present 100

percent authorized rated thermal power

level of 3579 megawatts thermal to 3758

megawatts thermal. This represents a

power level increase of 5 percent for the

Perry Nuclear Power Plant.

Date of issuance: June 1, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 112.

Facility Operating License No. NPF–

58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR

59802)

The supplemental information

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 1, 2000.

No significant hazards consideration

comments received: No.

IES Utilities Inc., Docket No. 50–331,

Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment:

February 18, 1999, as supplemented

September 15, 1999, and March 16,

2000.

Brief description of amendment: The

amendment revises Duane Arnold

Energy Center (DAEC) Technical

Specification (TS) Table 3.3.6.1–1,

‘‘Primary Containment Isolation

Instrumentation,’’ by deleting the

manual initiation function of the high

pressure coolant injection (HPCI) system

and reactor core isolation cooling (RCIC)

system isolation. A related condition as

well as corresponding surveillance

requirements and bases are also deleted.

Date of issuance: June 1, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 231.

Facility Operating License No. DPR–

49: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 7, 1999 (64 FR 17026).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 1, 2000.

No significant hazards consideration

comments received: No.

North Atlantic Energy Service

Corporation, et al., Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

December 16, 1998.

Description of amendment request:

The amendment makes several editorial

and administrative changes to the

following sections of the Technical

Specifications (TSs), Index Page vi,

‘‘Figures 3.4–2 and 3.4–3’’; Index Page

xv, ‘‘6.0 Administrative Controls’’;

4.2.4.2b, ‘‘Determination of Quadrant

Power Tilt Ratio’’; 6.4.1.7b, ‘‘SORC

Responsibilities’’; 6.4.2.2d, ‘‘Station

Qualified Reviewer Program’’; 6.3.1,

‘‘Training’’; 6.4.3.9c, ‘‘Records of

NSARC’’; 6.8.1.6.b.1, ‘‘Core Operating

Limits Report’’; and 6.8.1.6.b.10, ‘‘Core

Operating Limits Report’’. In addition,

the following Bases sections have been

revised: Bases 2.2.1, ‘‘Reactor Trip

System Instrumentation Setpoints’’;

Bases 3/4.2.4, ‘‘Quadrant Power Tilt

Ratio’’; Bases 3/4.2.5, ‘‘DNB

Parameters’’; Bases 3/4.4.8, ‘‘Specific

Activity’’; and Bases 3/4.5.1,

‘‘Accumulators’’.

Date of issuance: May 22, 2000.

Effective date: As of its date of

issuance, and shall be implemented

within 90 days.

37434 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

Amendment No.: 70.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 10, 1999 (64 FR

6700).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 22, 2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

April 12, 2000.

Brief description of amendment: The

amendment corrects a reference in

Technical Specification Section

6.9.1.8b.1, ‘‘Core Operating Limits

Report.’’

Date of issuance: May 26, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 246.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 21, 2000 (65 FR 21486).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 26, 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:

May 19, 2000.

Brief description of amendments: The

amendments delay implementation of

the improved Technical Specifications

to June 30, 2000 from May 31, 2000.

Date of issuance: May 24, 2000.

Effective date: May 24, 2000.

Amendment Nos.: Unit 1—141; Unit

2—141.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised Appendix D of the Operating

Licenses.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendments, finding of

emergency circumstances, and final

determination of no significant hazards

consideration are contained in a Safety

Evaluation dated May 24, 2000.

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

PECO Energy Company, Docket Nos.

50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

May 26, 1999.

Brief description of amendments: The

amendments remove Technical

Specification (TS) Surveillance

Requirement 4.1.3.5.b, control rod

scram accumulators’ alarm

instrumentation, and relocate it to the

Updated Final Safety Analysis Report

and plant procedures; and revise TS

Action Statement 3.1.3.5.a.2.a to allow

for an alternate method of determining

whether a control rod drive pump is

operating.

Date of issuance: May 22, 2000.

Effective date: The amendments are

effective as of the date of their issuance

and shall be implemented within 30

days. In addition, the licensee shall

include the relocated information in the

Updated Final Safety Analysis Report

submitted to the NRC, pursuant to 10

CFR 50.71(e), as was described in the

licensee’s application dated May 26,

1999 and evaluated in the staff’s safety

evaluation dated May 22, 2000.

Amendment Nos.: 143 and 105.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15382).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 22, 2000.

No significant hazards consideration

comments received: No.

Portland General Electric Company, et

al., Docket No. 50–344, Trojan Nuclear

Plant, Columbia County, Oregon

Date of application for amendment:

August 27, 1998, as supplemented by

letter dated July 1, 1999.

Brief description of amendment: The

amendment revises the Permanently

Defueled Technical Specifications to

delete the requirement for defueled

emergency plan procedures. This

amendment is contingent upon the

transfer of the nuclear spent fuel from

the existing 10 CFR Part 50 licensed

area to the 10 CFR Part 72 independent

spent fuel storage installation area.

Date of issuance: May 10, 2000.

Effective date: May 10, 2000, and

shall be implemented within 30 days

after the transfer of the last cask of spent

nuclear fuel from the spent fuel pool to

the independent spent fuel storage

installation is complete.

Amendment No.: 202.

Facility Operating License No. NPF–1:

The amendment changes the

Permanently Defueled Technical

Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR

46441).

The July 1, 1999, supplemental letter

provided additional clarifying

information and did not expand the

scope of the application as originally

noticed and did not change the staff’s

original no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 10, 2000.

No significant hazards consideration

comments received: No.

PP&L, Inc., Docket Nos. 50–387 and 50–

388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

March 14, 2000, as supplemented March

27, and May 25, 2000.

Brief description of amendments: The

amendments extended the

implementation date for Amendment

No. 184 to Facility Operating License

NPF–14 and Amendment No. 158 to

Facility Operating License NPF–22 from

30 days following startup from the Unit

1 Spring 2000 refueling outage to no

later than November 1, 2001.

Date of issuance: June 2, 2000.

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment Nos.: 187 and 161.

Facility Operating License Nos. NPF–

14 and NPF–22. The amendments

revised the license.

Date of initial notice in Federal

Register: April 27, 2000 (65 FR 24718).

The May 25, 2000, letter provided

clarifying information but did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 2, 2000.

No significant hazards consideration

comments received: No.

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of application for amendment:

May 3, 2000, as supplemented on May

19, 2000.

Brief description of amendment: The

license amendment modifies the

existing requirement under Technical

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37435

Specification Section 3.1.3.2.1, Action

a.1, to determine the position of Rod

1SB2 from once every 8 hours to within

8 hours following any movement of the

rod until repair of the rod indication

system is completed. This change is

applicable for the remainder of the Unit

1 Cycle 14, or until an outage of

sufficient duration occurs whereby the

licensee can repair the position

indication system.

Date of issuance: May 26, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 30 days.

Amendment No.: 230

Facility Operating License No. DPR–

70: This amendment revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (65 FR

30137) May 10, 2000. The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. That

notice also provided for an opportunity

to request a hearing by May 24, 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 is contained in a

Safety Evaluation dated May 26, 2000.

South Carolina Electric & Gas

Company, South Carolina Public

Service Authority, Docket No. 50–395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of application for amendment:

January 27, 2000.

Brief description of amendment: The

amendment revises the spent fuel pool

reactivity limit requirement by

removing the value for K infinity from

Specification 5.6.1.1 and replacing it

with a figure of integral fuel burnable

absorbers rods versus nominal Uranium235 enrichment.

Date of issuance: June 1, 2000.

Effective date: June 1, 2000.

Amendment No.: 144.

Facility Operating License No. NPF–

12: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR

9011).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 1, 2000.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–260, and 50–296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone

County, Alabama

Date of application for amendments:

March 15, 2000.

Description of amendment request:

The amendments revised the Technical

Specifications (TS) to provide a 7-day

limiting condition for operation when

two trains of the Containment Air

Dilution System are inoperable.

Date of issuance: May 24, 2000.

Effective date: May 24, 2000.

Amendment Nos.: 265 and 225.

Facility Operating License Nos. DPR–

52 and DPR–68. Amendments revised

the TS.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17919).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 24, 2000.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

April 29, 1999.

Brief description of amendments:

These amendments revise Technical

Specification (TS) Section 3/4.3.3,

‘‘Radiation Monitoring

Instrumentation,’’ TS Section 3/4.7.7,

‘‘Control Room Emergency Ventilation

System,’’ and the associated bases.

Actions are added and modified

regarding inoperable equipment.

Date of issuance: May 31, 2000.

Effective date: May 31, 2000.

Amendment Nos.: 256 and 247.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: May 19, 1999 (64 FR 27325).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 31, 2000.

No significant hazards consideration

comments received: No.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

March 6, 2000 (ULNRC–04197).

Brief description of amendment: The

amendment revises Limiting Condition

for Operation (LCO) 3.7.1, ‘‘Main Steam

Safety Valves (MSSVs),’’ in that the

maximum allowable reactor power for a

given number of operable MSSVs per

steam generator is reduced in Table

3.7.1–1, ‘‘Operable Main Steam Safety

Valves [MSSVs] versus Maximum

Allowable Power,’’ and in Required

Action A.1 of the TSs. These changes

will result in decreasing the setpoint

values for the power range neutron flux

high channels, which are part of the

reactor trip system (RTS)

instrumentation in Table 3.3.1–1,

‘‘Reactor Trip System Instrumentation,’’

and will result in the reactor operating

at a lower power for a given number of

operable MSSVs per steam generator. In

addition, two format errors in the

actions for LCO 3.7.1 are corrected.

Date of issuance: May 26, 2000.

Effective date: May 26, 2000, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 136.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17920).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 26, 2000.

No significant hazards consideration

comments received: No.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

37436 Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective 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 application for

amendment, (2) the amendment to

Facility Operating License, 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).

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By July

14, 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

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. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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:

Rulemakings and Adjudications Staff or

may be delivered to the Commission’s

Public Document Room, the Gelman

Federal Register / Vol. 65, No. 115 / Wednesday, June 14, 2000 / Notices 37437

1The radioactive materials, sometimes referred to

as agreement materials, are: (a) Byproduct materials

as defined in Section 11e.(1) of the Act; (b)

byproduct materials as defined in Section 11e.(2) of

the Act; (c) source materials as defined in Section

11z. of the Act; and (d) special nuclear materials as

defined in Section 11a. of the Act, restricted to

quantities not sufficient to form a critical mass.

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 the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: May 12,

2000, as supplemented by letter dated

May 19, 2000.

Brief description of amendment: The

amendment revises TS 3.7.3, Condition

A, to extend the Completion Time for

one or more feedwater isolation valves

(FIVs) inoperable from 4 hours to 24

hours if, within 4 hours, the respective

feedwater control valves (FCVs) and the

FCV bypass valves in the same flowpath

are verified to be capable of performing

the feedwater isolation function. A

footnote is added that indicates that the

extension of the Completion Time to 24

hours is only applicable for repair of the

FIV hydraulic system through fuel cycle

8 for Unit 1 and fuel cycle 5 for Unit 2.

Date of issuance: May 25, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 77.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendment revises

the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes The NRC published

a public notice of the proposed

amendment, issued a proposed finding

of no significant hazards consideration

and requested that any comments on the

proposed no significant hazards

consideration be provided to the staff by

the close of business on May 24, 2000.

The notice was published in the Dallas

Morning News and the Ft. Worth Star

Telegram from May 21 through May 23,

2000.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, consultation with the

State of Texas, and final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

May 25, 2000.

Dated at Rockville, Maryland, this 7th day

of June 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–14837 Filed 6–13–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

State of Oklahoma: NRC Staff

Assessment of a Proposed Agreement

Between the Nuclear Regulatory

Commission and the State of

Oklahoma

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of a proposed Agreement

with the State of Oklahoma.

SUMMARY: This notice is announcing

that the Nuclear Regulatory Commission

(NRC) has received a request from

Governor Frank Keating of Oklahoma

that the NRC consider entering into an

Agreement with the State as authorized

by Section 274 of the Atomic Energy Act

of 1954, as amended (Act). Section 274

of the Act contains provisions for the

Commission to enter into agreements

with the Governor of any State

providing for the discontinuance of the

regulatory authority of the Commission.

Under the proposed Agreement,

submitted December 28, 1999, the

Commission would discontinue and

Oklahoma would take over portions of

the Commission’s regulatory authority

over radioactive material covered under

the Act within the State of Oklahoma.

In accordance with 10 CFR 150.10,

persons, who possess or use certain

radioactive materials in Oklahoma,

would be released (exempted) from

portions of the Commission’s regulatory

authority under the proposed

Agreement. The Act requires that NRC

publish those exemptions. Notice is

hereby given that the pertinent

exemptions have been previously

published in the Federal Register and

are codified in the Commission’s

regulations as 10 CFR Part 150. NRC is

publishing the proposed Agreement for

public comment, as required by the Act.

NRC is also publishing the summary of

an assessment conducted by the NRC

staff of the proposed Oklahoma

byproduct material regulatory program.

Comments are invited on (a) the

proposed Agreement, especially its

effect on public health and safety, and

(b) the NRC staff assessment.

DATES: The comment period expires July

7, 2000. Comments received after this

date will be considered if it is practical

to do so, but the Commission cannot

assure consideration of comments

received after the expiration date.

ADDRESSES: Written comments may be

submitted to Mr. David L. Meyer, Chief,

Rules and Directives Branch, Division of

Administrative Services, Office of

Administration, Washington, DC 20555–

0001. Copies of comments received by

NRC may be examined at the NRC

Public Document Room, 2120 L Street,

NW. (Lower Level), Washington, DC.

Copies of the proposed Agreement,

copies of the request for an Agreement

by the Governor of Oklahoma including

all information and documentation

submitted in support of the request, and

copies of the full text of the NRC staff

assessment are also available for public

inspection in the NRC’s Public

Document Room.

FOR FURTHER INFORMATION CONTACT:

Patricia M. Larkins, Office of State and

Tribal Programs, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001. Telephone (301) 415–

2309 or e-mail pml@nrc.gov.

SUPPLEMENTARY INFORMATION: Since

Section 274 of the Act was added in

1959, the Commission has entered into

Agreements with 31 States. The

Agreement States currently regulate

approximately 16,000 agreement

material licenses, while NRC regulates

approximately 5800 licenses. Under the

proposed Agreement, approximately

220 NRC licenses will transfer to

Oklahoma. NRC periodically reviews

the performance of the Agreement States

to assure compliance with the

provisions of Section 274. Section 274e

requires that the terms of the proposed

Agreement be published in the Federal

Register for public comment once each

week for four consecutive weeks. This

notice is being published in fulfillment

of the requirement.

I. Background

(a) Section 274d of the Act provides

the mechanism for a State to assume

regulatory authority, from the NRC, over

certain radioactive materials 1 and

activities that involve use of the

materials. In a letter dated December 28,