69058 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

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, DC 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 wmn@nrc.gov or

dkw@nrc.gov.

Dated: November 9, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–29354 Filed 11–13–00; 2:18 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 October 23,

2000, through November 3, 2000. The

last biweekly notice was published on

November 1, 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.

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 and

Directives Branch, Division of

Administrative 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’s Public

Document Room, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852. The

filing of requests for a hearing and

petitions for leave to intervene is

discussed below.

By December 15, 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, located at One

White Flint North, 11555 Rockville Pike

(first Floor), Rockville, Maryland 20852.

Publicly available records will be

accessible 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

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69059

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, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852, 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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly 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).

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request:

November 22, 1999, as supplemented on

September 11, 2000.

Description of amendment request:

The proposed amendment would revise

Technical Specification Sections 4.5.D,

‘‘Containment Air Filtration System

(CAFS),’’ 4.5.E, ‘‘Control Room Air

Filtration System (CRAFS),’’ 4.5.F,

‘‘Fuel Storage Building Air Filtration

System (FSBAFS),’’ and 4.5.G, ‘‘Postaccident Containment Venting System

(PACVS),’’ to address the testing

requirements in Generic Letter 99–02,

‘‘Laboratory Testing of Nuclear-Grade

Activated Charcoal.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed change would revise

Section 4.5 to incorporate current NRC

[Nuclear Regulatory Commission] testing

requirements which affect how the charcoal

would be tested in the laboratory. These

changes would not affect possible initiating

events for accidents previously evaluated or

alter the configuration or operation of the

facility. The Limiting Safety System Settings

and Safety Limits specified in the current

Technical Specifications would remain

unchanged. Therefore, the proposed changes

would not involve a significant increase in

the probability or in the consequences of an

accident previously evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed changes would

implement testing methodology for

ventilation system charcoal in accordance

with Generic Letter 99–02, but would not

alter equipment performance criteria or

standards. The safety analysis of the facility

would remain complete and accurate, and

would not be affected by the new charcoal

testing requirements. There would be no

physical changes to the facility and the plant

conditions for which the design basis

accidents have been evaluated would still be

valid. The operating procedures and

emergency procedures would be unaffected.

Consequently no new failure modes would

be introduced as a result of the proposed

change. Therefore, the proposed changes

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. Since there would be no changes to the

operation of the facility, to its physical

design, or to the performance characteristics

of any safety-related equipment, neither the

Updated Final Safety Analysis Report

(UFSAR) design basis, accident assumptions,

nor Technical Specification bases would be

affected. 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: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Section Chief: Marsha

Gamberoni.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of amendment request:

September 5, 2000.

Description of amendment request:

The amendment revises Technical

Specification 3.3.5.1, 3.3.6.1 and 3.3.6.2.

The proposed changes would add notes

to tables listing instrument channels

that are common to, or support the

operability of interrelated systems as

governed by these technical

specifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

69060 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change has no impact on

previously analyzed accidents or transients

and has no affect on design, operation,

capacity, or surveillance requirements of the

affected instrumentation channels. The

change provides branching notes to the Loss

of Coolant Accident (LOCA) Time Delay

Relay (TDR) Functions of LCO [limiting

condition of operation] 3.3.5.1 from

instrument channels of the primary and

secondary containment isolation channels of

LCO 3.3.6.1 and LCO 3.3.6.2 and the

associated support features for the LOCA

TDR function. Since these instruments affect

multiple LCOs, this change will assure that

operators implement the most restrictive

Action and Completion Time when a channel

becomes inoperable or is placed in the

tripped condition. Providing this branching

to the more restrictive Actions makes explicit

what is currently required for Operability

and has no impact on any previously

evaluated accident.

Therefore, operation of WNP–2 in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not impact any

operational or physical aspect of WNP–2.

The change only makes explicit the LCOs

affected by the primary and secondary

containment isolation instruments and the

associated supported features for the LOCA

TDR function.

Therefore, 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.

The proposed change provides branching

notes to the LOCA TDR channels of LCO

3.3.5.1 from instrument channels of the

primary and secondary containment isolation

channels of LCO 3.3.6.1 and LCO 3.3.6.2 and

provides notes for identifying associated

support features for the LOCA TDR function.

This change only makes explicit what is

currently required for LCO 3.3.5.1 Functions

1c, 1d, 2c and 2d instrument channel

Operability. This change will make explicit

the most restrictive Action when an

instrument sensor or channel becomes

inoperable or is placed in the tripped

condition, thereby, maintaining the margin of

safety in accordance with the Technical

Specifications.

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.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: October

3, 2000.

Description of amendment request:

The proposed amendment would revise

the Crystal River Unit 3 (CR–3)

Improved Technical Specifications (ITS)

3.7.12, ‘‘Control Room Emergency

Ventilation System (CREVS),’’ ITS

5.6.2.12, ‘‘Ventilation Filter Testing

Program (VFTP),’’ ITS 3.3.16, ‘‘Control

Room Isolation—High Radiation,’’ and

ITS 3.7.18, ‘‘Control Complex Cooling

System.’’ The proposed ITS changes are

based on the results of revised public

and control room dose calculations for

CR–3 design basis radiological accidents

using an alternative source term (AST).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

1. Does not involve a significant increase

in the probability or consequences of an

accident previously analyzed.

The proposed amendment does not involve

a significant increase in the probability or

consequences of an accident previously

analyzed. The CR–3 Control Room

Emergency Ventilation System (CREVS) and

the Control Complex Habitability Envelope

(CCHE) only function following the initiation

of a design basis radiological accident.

Therefore, the changes to the CREVS

specification, the CREVS filter testing

criteria, and the deletion of the requirement

for control room isolation on high radiation

proposed by this amendment will not

increase the probability of any previously

analyzed accident. The Control Complex

Cooling System and Auxiliary Building

Ventilation System are not initiators of any

design basis accident. Therefore, the changes

to the Control Complex Cooling System

specification and the changes to the testing

guidelines for the Auxiliary Building

Ventilation System exhaust filters proposed

by this amendment will not increase the

probability of occurrence of any previously

analyzed accident.

Revised dose calculations, which take into

account the changes proposed by this

amendment and the use of an AST, have

been performed for the CR–3 design basis

radiological accidents. The results of these

revised calculations indicate that public and

control room doses will not exceed the limits

specified by 10 CFR 50.67 and Regulatory

Guide 1.183. In addition, a comparison

between results of the current public dose

calculations and the revised public dose

calculations indicate that the proposed

changes will not result in a significant

increase in predicted dose consequences for

any of the analyzed accidents. Therefore, the

proposed changes do not involve a

significant increase in the consequences of

any previously analyzed accident.

2. Does not create the possibility of a new

or different kind of accident from any

accident previously analyzed.

Limiting the requirements for the Control

Complex Cooling System and CREVS to be

operable to Modes 1, 2, 3, and 4, and

changing the Auxiliary Building Ventilation

System exhaust filter testing guidelines do

not result in changes to the design or

operation of these systems. Although the

other changes proposed by this amendment

could affect the operation of the CREVS and

CCHE following a design basis radiological

accident, none of these changes can initiate

a new or different kind of accident since they

are only related to system capabilities that

provide protection from accidents that have

already occurred. Therefore the proposed

changes do not create the possibility of a new

or different kind of accident from those

previously analyzed.

3. Does not involve a significant reduction

in the margin of safety.

The proposed changes to the control

complex cooling specification do not affect

the ability of the system to maintain control

complex temperatures within safety-related

equipment operability limits when the

equipment is required. The results of revised

control room dose calculations indicate that

the proposed changes to the CREVS

specification, the CREVS filter testing

criteria, and removal of the CREVS actuation

signal on high radiation will not affect the

ability of the CREVS and CCHE to maintain

control room doses less than required limits

during design basis radiological accidents.

The revised dose calculations also indicate

that the Auxiliary Building Ventilation

System exhaust filters are not required in

order to maintain public or control room

doses less than required limits; therefore the

proposed changes to the testing requirements

for these filters cannot adversely affect public

or control room doses.

Based on the above, the revised technical

specifications meet the same intent as the

currently approved specifications. Therefore,

the proposed changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC—A5A, P.O. Box

14042, St. Petersburg, Florida 33733–

4042.

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69061

NRC Section Chief: Richard P.

Correia.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: June 7,

1999, as supplemented February 4,

2000.

Description of amendment request:

The proposed amendment requests the

staff to evaluate the integrity of the

Kewaunee Reactor Pressure Vessel

(RPV) circumferential beltline weld

using a Master Curve-based

methodology.

The licensee submitted a request for

exemptions to 10 CFR 50.61, 10 CFR 50

Appendix G, and 10 CFR 50, Appendix

H, to allow the use of the Master Curvebased methodology for calculating the

RPV Reference Temperature for

Pressurized Thermal Shock (RTPTS)

based on the fracture toughness data

from irradiated pre-cracked Charpy Vnotch specimen testing of Kewaunee

and Maine Yankee surveillance welds.

The Master Curve methodology is based

on American Society for Mechanical

Engineers (ASME) Code Case N–629 and

American Society for Testing and

Materials Standard (ASTM) E–1921. In

its submittals, the licensee also

requested a revision of the facility’s

Pressure-Temperature (P/T) limit

curves.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Failure of a reactor vessel is not an

accident that has been previously evaluated.

Design provisions ensure that this is not a

credible event. Since the potential

consequences of a reactor vessel failure are

so severe, industry and governmental

agencies have worked together to ensure that

failure will not occur. Compliance with 10

CFR 50.61, 10 CFR 50 Appendix G and H,

and application of ASME Code Case N–514,

ASME Code Case N–588, and the exemption

requested in Attachment 1 ensures that

failure of a reactor vessel will not occur. The

proposed changes do not impact the

capability of the reactor coolant pressure

boundary piping (i.e., no change in operating

pressure, materials, seismic loading, etc.) and

therefore do not increase the potential for the

occurrence of a LOCA.

The LTOP setpoint, LTOP system enabling

temperature, and revised P/T limits reflected

in proposed Figures TS 3.1–1 and TS 3.1–2

ensure that the Appendix G pressure/

temperature limits are not exceeded, and

therefore, ensure that RCS integrity is

maintained. The changes do not modify the

reactor coolant system pressure boundary,

nor make any physical changes to the facility

design, material, construction standards, or

setpoints. The reactor coolant system full

power operating pressure (2235 psig) is not

being changed by this proposed amendment.

The LTOP valve setpoint remains at ≤500

psig. The LTOP enabling temperature based

on Figure TS 3.1–2 is 200°F and is consistent

with ASME Code Case N–514 guidance of

RTNDT + 50°F. The LTOP enabling

temperature is not changed by this

amendment. The allowable combination of

Appendix G pressure and temperature for the

cooldown limits is marginally greater than

the current limits. The combination of

slightly greater allowable Appendix G

pressure and temperature limits and low

enabling temperature produces an adequate

operating window. An adequate operating

window reduces the likelihood of

inadvertently lifting the LTOP relief valve

while maneuvering the plant through the

knee of the P–T curve during startup and

shutdown. The probability of an LTOP event

occurring is independent of the pressuretemperature limits for the RCS pressure

boundary and enabling temperature.

Therefore, the probability of a LTOP event is

not increased.

The revised heatup and cooldown limit

curves and corresponding LTOP enabling

temperature were developed using test

results from unirradiated and/or irradiated

specimens that represent the KNPP reactor

vessel beltline circumferential weld, closure

head flange, and intermediate forging. The

circumferential beltline weld and

intermediate forging are the most limiting

materials in the reactor coolant pressure

boundary. These materials are limiting due to

the effects of neutron irradiation which cause

the flow properties to increase and the

toughness to decrease. The circumferential

beltline weld is the controlling material for

evaluation of pressurized thermal shock.

With NRC approval to use Code Case N–588

and the exemption requested in Attachment

1, the reactor vessel intermediate forging and

head flange become the limiting and

controlling materials for development of the

Appendix G limit curves and corresponding

LTOP system enabling temperature. 10 CFR

50, Appendix G states that the metal

temperature of the closure flange regions

must exceed the material unirradiated RTNDT

by at least 120°F for normal operation and

90°F for hydrostatic pressure tests and leak

tests when the pressure exceeds 20 percent

of the preservice hydrostatic test pressure.

Fracture toughness, drop weight, and Charpy

V-notch testing of the 1P3571 weld metal and

drop weight, and Charpy V-notch testing of

the intermediate forging material has been

performed. The results of those tests have

been used for derivation of the revised PTS

assessment, the proposed Appendix G heatup

and cooldown limit curves, and the

corresponding LTOP system enabling

temperature. The revised limit curves and

corresponding LTOP enabling temperature

have been developed using accepted

engineering practices. The evaluations were

performed in accordance with methods

derived from the ASME Boiler and Pressure

Vessel Code, criteria set forth in NRC

Regulatory Standard Review Plan 5.3.2, and

10 CFR 50.61. The revised heatup and

cooldown limit curves and corresponding

LTOP enabling temperature ensures adequate

fracture toughness for ferritic materials of the

pressure-retaining components of the reactor

coolant pressure boundary. These limit

curves provide adequate margins of safety

during any condition of normal operation,

including anticipated operational

occurrences and system hydrostatic tests, and

low temperature overpressure protection

[corresponding to isothermal events during

low temperature operations (i.e., ≤200°F)],

thus ensuring the integrity of the reactor

coolant pressure boundary.

The changes do not adversely affect the

integrity of the RCS such that its function in

the control of radiological consequences is

affected. Radiological off-site exposures from

normal operation and operational transients,

and faults of moderate frequency do not

exceed the guidelines of 10 CFR 100. In

addition, the changes do not affect any

fission product barrier. The changes do not

degrade or prevent the response of the LTOP

relief valve or other safety-related systems to

previously evaluated accidents. In addition,

the changes do not alter any assumption

previously made in the radiological

consequence evaluations nor affect the

mitigation of the radiological consequences

of an accident previously evaluated.

Therefore, the consequences of an accident

previously evaluated will not be increased.

Thus, operation of KNPP in accordance

with the PA [proposed amendment] does not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Since the potential consequences of a

reactor vessel failure are so severe, industry

and governmental agencies have worked

together to ensure that failure will not occur.

Application of ASME Code Case N–514,

ASME Code Case N–588, and the exemption

requested in Attachment 1 ensures that

failure of a reactor vessel will not occur.

Therefore, a failure of the reactor vessel can

still be considered incredible.

The proposed heatup and cooldown limit

curves have been constructed by combining

the most conservative pressure-temperature

limits derived by using material properties of

the intermediate forging, closure head flange,

and beltline circumferential weld to form a

single set of composite curves. Use of the

proposed curves, does not modify the reactor

coolant system pressure boundary, nor make

any physical changes to the LTOP setpoint or

design. Proposed Figures TS 3.1–1 and TS

3.1–2 were prepared in accordance with

regulatory and code requirements and were

derived using conservative material property

basis and neutron exposure projections thru

33 EFPY. Therefore, the proposed heatup and

cooldown curves and LTOP limits will

continue to protect the reactor vessel from

failure.

The LTOP system enabling temperature

and the proposed Appendix G pressure

69062 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

temperature limitations were prepared using

methods derived from the ASME Boiler and

Pressure Vessel Code and the criteria set

forth in NRC Regulatory Standard Review

Plan 5.3.2. The changes do not cause the

initiation of any accident nor create any new

credible limiting failure for safety-related

systems and components. The changes do not

result in any event previously deemed

incredible being made credible. As such, it

does not create the possibility of an accident

different than previously evaluated. The

changes do not have any adverse effect on the

ability of the safety-related systems to

perform their intended safety functions.

The proposed changes do not make

physical changes to the plant or create new

failure modes. Thus, the PA [proposed

amendment] does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed Appendix G pressure

temperature limitations and corresponding

LTOP enabling temperature were prepared

using methods derived from the ASME Boiler

and Pressure Vessel Code, including ASME

Code Cases N–514 , N–588, and N–629.

Inherent conservatism in the P/T limits

resulting from these documents is described

in the Safety Evaluation.

Alternative methodologies to the safety

margins required by Appendix G to 10 CFR

Part 50 have been developed by the ASME

Working Group on Operating Plant Criteria.

Three of these methodologies are contained

in ASME Code Cases N–514, N–588, and N–

629.

Code Case N–514 provides criteria to

determine pressure limits during LTOP

events that avoid certain unnecessary

operational restrictions, provide adequate

margins against failure of the reactor pressure

vessel, and reduce the potential for

unnecessary activation of the relief valve

used for LTOP. Specifically, the ASME Code

Case N–514 allows determination of the

setpoint for LTOP events such that the

maximum pressure in the vessel would not

exceed 110% of the P/T limits of the existing

ASME Appendix G; and redefines the

enabling temperature at a coolant

temperature less than 200 °F or a reactor

vessel metal temperature less than RTNDT +

50 °F, whichever is greater. Code Case N–

514, ‘‘Low Temperature Overpressure

Protection,’’ has been approved by the ASME

Code Committee but not yet approved for use

in Regulatory Guides 1.147, 1.85, or 1.84. The

content of this Code Case has been

incorporated into Appendix G of Section XI

of the ASME Code and published in the 1993

Addenda to Section XI. It is expected that the

next revision of 10 CFR 50.55a will endorse

the 1993 Addenda and Appendix G of

Section XI. Code Case N–514 is not in

conflict with 10 CFR 50.61 and therefore has

been used to establish the LTOP system

enabling temperature; the provision for

exceeding 110% of the Appendix G limits

has not been incorporated in PA [proposed

amendment] 160. The NRC previously

approved use of Code Case N–514 for

determination of the LTOP enabling

temperature in Reference 6.

Code Case N–588 provides benefits in

terms of calculating pressure-temperature

limits by revising the Section XI, Appendix

G reference flaw orientation for

circumferential welds in reactor vessels. The

NRC previously approved use of Code Case

N–588 for use at KNPP in references 4 and

5.

In support of this PA [proposed

amendment], WPSC used fracture toughness

results representing the beltline weld metal

that were irradiated to EOL and in excess of

EOLE fluence. The fracture toughness results

were analyzed as described under Case #6 in

WCAP–15075 and ASME Code Case N–629

for determining the EOL and EOLE indexing

reference temperature values. Attachment 1

to this letter provides information to support

NRC approval to use the weld metal fracture

toughness results along with the

methodology presented in WCAP–15075 for

the KNPP PTS evaluation. The KNPP

application of the methodology presented in

WCAP–15075, identified as Case #6,

incorporates the following additional

margins beyond that recommended in ASTM

E1921–97:

(a) A delta value of 17 °F is added to T0

to ensure that the margin in the KNPP

application is at least as conservative as the

margin associated with the most limiting

HSST–02 plate material.

(b) An additional margin of 18 °F has been

added to the above 17 °F to be consistent

with the ASME Code Case N–629, and align

the KNPP lead plant application with current

consensus of the technical community

regarding the best use of fracture toughness

based indexing reference temperature data.

(c) A 2 σ value of 16 °F and 24 °F is added

to account for RTTo measurement uncertainty

for EOL and EOLE, respectively.

(d) A value of (+)35 °F and (¥)32 °F

accounts for heat uncertainty between the

KNPP and Maine Yankee surveillance

capsule specimens for EOL and EOLE,

respectively.

Fracture toughness testing of irradiated

1P3571 weld metal, performed in accordance

with ASTM E1921–97 and application of

ASME Code Case N–629 along with the

methods in WCAP–15075, indicate that the

end of life indexing reference temperature is

234 °F. This fracture toughness generated

EOL indexing reference temperature value

includes a margin of 34 °F (18 °F + 16 °F).

The fracture toughness generated indexing

reference temperature value (234 °F) is lower

than the ART value (277 °F) predicted by the

Charpy V-notch and Drop Weight

methodology. Both methodologies predict

end of life indexing reference temperature

values that are below the pressurized thermal

shock screening criteria (300 °F).

Use of the methodology set forth in the

ASME Boiler and Pressure Vessel Code, NRC

Regulatory Standard Review Plan 5.3.2.,

WCAP–15075, 10 CFR 50.61, and 10 CFR 50

Appendices G and H ensures that proper

limits and safety factors are maintained.

Thus, the PA [proposed amendment] does

not involve a significant reduction in the

margin of safety.

The revised heatup and cooldown limit

curves and corresponding LTOP system

enabling temperature were prepared using

fracture toughness, drop weight and Charpy

V-notch data for the beltline weld material;

drop weight and Charpy V-notch data for the

closure head flange and intermediated

forging material; along with practices

described herein and methods derived from

the ASME Boiler and Pressure Vessel Code

and 10 CFR 50.61. The safety factors and

margins used in the development of the limit

curves and LTOP system enabling

temperature meet the criteria set forth by

these documents. Application of low leakage

core designs decreases the rate of shift in

transition temperature from ductile to

nonductile behavior. The revised limit curves

and corresponding LTOP enabling

temperature provide adequate margins of

safety during any condition of normal

operation, including anticipated operational

occurrences and system hydrostatic tests, and

low temperature overpressure protection

[corresponding to isothermal events during

low temperature operations (i.e., ≤200 °F)].

With the preparation of the revised limit

curves in accordance with the latest criteria

and guidance, this proposed amendment

ensures that proper limits and safety factors

are maintained.

Thus, the proposed amendment does not

involve a significant reduction in a margin of

safety. Therefore, the proposed amendment

does not represent a significant decrease in

the margin of safety. As shown in

Attachment 1 [in the proposed amendment],

a loss of reactor vessel integrity is still

incredible. Furthermore, the LTOP setpoint

and enabling temperature will continue to

protect the reactor coolant system during low

temperature operation.

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: Bradley D.

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

Box 1497, Madison, WI 53701–1497.

NRC Section Chief: Claudia M. Craig.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request:

September 7, 2000.

Description of amendment request:

The proposed amendment to the Indian

Point Nuclear Generating Unit No. 3

(IP3) Technical Specifications (TSs)

would reflect a modification planned for

refueling outage (RO) 11, scheduled to

begin in May of 2001. The modification

will automatically close, on a safety

injection signal, the existing main

feedwater inlet isolation valves (MFIIVs)

and the main feedwater low flow bypass

inlet isolation valves.

Basis for proposed no significant

hazards consideration determination:

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69063

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 Indian Point 3 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:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed TS change reflects a planned

modification to automatically isolate main

feedwater on a safety injection signal using

the motor operated Main Feedwater Inlet

Isolation Valves (MFIIVs) and MF [main

feedwater] low flow bypass inlet isolation

valves. These non-safety valves will be

incorporated into the IST [inservice testing]

program as augmented components and

included in the Generic Letter 89–10 program

for motor operated valves. The modification

will not relocate the safety injection signal

from the Main Boiler Feedpump Discharge

Valves (MBFPDVs) but closure will no longer

be assumed in analyses. The modification is

based on current design function for the

feedwater isolation following a main steam

line break inside containment accomplished

by MBFPDVs. The TS changes add a limiting

condition for operation, required action

statements with completion times and

surveillance requirements that are the same

as those previously approved for

Westinghouse plants in the Standard

Technical Specifications found in NUREG–

1432. The plant core reload analysis will

assume that the modification is complete

(this eliminates the continued addition of the

feedwater between the MFIIVS and

associated bypass valves and the MBFPDVs)

and demonstrate that a shutdown margin of

1.3% is acceptable and that no boron

concentration needs to be assumed in the

safety injection lines. The proposed changes

cannot affect the probability of an accident

occurring since they reflect a change in plant

design consistent with current design which

is not an accident initiator. The proposed

changes cannot increase the consequences of

postulated accidents since they reflect a

change in plant design that will mitigate the

effects of feedwater to a faulted steam

generator for a main steam line break inside

containment and restore past analytical

assumptions regarding a 1.3% shutdown

margin and no boron in the safety injection

lines.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed TS change reflects a planned

modification to automatically isolate main

feedwater on a safety injection signal using

the motor operated Main Feedwater Inlet

Isolation Valves (MFIIVs) and MF low flow

bypass inlet isolation valves. These nonsafety valves will be incorporated into the

IST program as augmented components and

included in the Generic Letter 89–10 program

for motor operated valves. The modification

will not relocate the safety injection signal

from the Main Boiler Feedpump Discharge

Valves (MBFPDVs) but closure will no longer

be assumed in analyses. The modification is

based on current design function for the

feedwater isolation following a main steam

line break inside containment accomplished

by MBFPDVs. The TS changes add a limiting

condition for operation, required action

statements with completion times and

surveillance requirements that are the same

as those previously approved for

Westinghouse plants in the Standard

Technical Specifications found in NUREG–

1432. The proposed TS changes do not create

the possibility of a new or different type of

accident from those previously evaluated

since they reflect a design change that will

accomplish the same feedwater isolation

function as previously done by the MBFPDVs

with no change to the manner in which the

feedwater system operates.

3. Involve a significant reduction in a

margin of safety.

The proposed TS change reflects a planned

modification to automatically isolate main

feedwater on a safety injection signal using

the motor operated Main Feedwater Inlet

Isolation Valves (MFIIVs) and MF low flow

bypass inlet isolation valves. These nonsafety valves will be incorporated into the

IST program as augmented components and

included in the Generic Letter 89–10 program

for motor operated valves. The modification

will not relocate the safety injection signal

from the Main Boiler Feedpump Discharge

Valves (MBFPDVs) but closure will no longer

be assumed in analyses. The modification is

based on current design function for the

feedwater isolation following a main steam

line break inside containment accomplished

by MBFPDVs. The TS changes add a limiting

condition for operation, required action

statements with completion times and

surveillance requirements that are the same

as those previously approved for

Westinghouse plants in the Standard

Technical Specifications found in NUREG–

1432. The plant core reload analysis will

assume that the modification is complete

(this eliminates the continued addition of the

feedwater between the MFIIVS and

associated bypass valves and the MBFPDVs)

and demonstrate that a shutdown margin of

1.3% is acceptable and that no boron

concentration needs to be assumed in the

safety injection lines. The proposed TS

change cannot involve a significant reduction

in the margin of safety since it is based upon

a modification that will restore the margin of

safety with respect to feedwater addition,

shutdown margin and core boration for a

main steam line break inside containment to

the previously analyzed condition. This

assumes that loading of the valves on the

emergency diesel generators will not affect

the emergency diesel generators margin.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Mr. David E.

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Section Chief: Marsha

Gamberoni.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request:

September 7, 2000.

Description of amendment request:

The proposed amendment to the Indian

Point Nuclear Generating Unit No. 3

(IP3) Technical Specifications (TSs)

would extend allowed outage times

(AOTs) on a one-time basis, before May

31, 2002, to allow for replacement of the

31 and 32 station batteries while the

plant is on line. The proposed

amendment also removes an expired

footnote regarding repairs to the 32

diesel fuel oil tank.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed License amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No. The proposed AOT extension does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. During the replacement

of the existing station batteries, a temporary

battery will provide the same function as the

Exide batteries being removed. Even though

this temporary battery will not meet seismic,

seismic interaction or security requirements,

due to its location on the 53-ft elevation of

the Turbine Building, it is qualified as safety

related in all other respects. The 125 VDC

EDS [electrical distribution system] is

normally supplied by the associated 480 VAC

bus through a Battery Charger. The essential

function of 31, 32 and 33 station battery is

to supply DC control power necessary to start

and load the associated EDG [emergency

diesel generator]. Once the EDGs are on line,

the 125 VDC EDS will be supplied via the

battery charger. However, the station

batteries have been sized to carry shutdown

loads for a period of two hours without

battery terminal voltage falling below its

minimum required voltage following a plant

trip that includes a loss of all AC power. This

provides additional assurance that the

critical DC loads are available in the event of

a loss of the battery charger. During the 10-

day AOT, when the temporary battery and

the associated battery charger are supporting

the 125 VDC bus, the ability of that ESF

[engineered safety feature] DC power panel to

mitigate an event/accident remains

unchanged except for its ability to cope with

a seismic, seismic interaction or security

event. However, the probability of these

69064 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

types of events concurrent with the 10-day

AOT is very small. During these types of

events, one ESF DC power panel may be

compromised, however IP3 has adequate 125

VDC power available in the form of two other

ESF train DC power panels to mitigate all

DBAs. The postulated loss of one ESF DC

power panel is bounded by the loss of an

entire ESF electrical train, a condition which

the plant is currently evaluated to withstand.

Based upon the above, the overall design,

function and operation of the 125 VDC EDS

and equipment has not been significantly

modified by the proposed changes. The

proposed changes do not affect accident

initiators or precursors, nor do they alter the

design assumptions for the systems or

components used to mitigate the

consequences of an accident as analyzed in

Chapter 14 of the IP3 USFAR [UFSAR]

[updated final safety analysis report], except

for one of the three trains of DC power. The

remaining DC power trains can mitigate a

DBA [design-basis accident]. Therefore, the

proposed one-time AOT extension TS

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Does the proposed License Amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

No. During the replacement of the existing

station batteries, a temporary battery will

provide the same function as the batteries

being removed. Even though this temporary

battery does not meet all design requirements

of a seismic, seismic interaction or security

event it possesses adequate capacity to fulfill

the safety related requirements of supplying

necessary power to the associated 125 VDC

bus under most conditions. Because the

temporary battery will perform like the

station battery that is currently installed, and

will be connected and used in the same way

as a backup power supply to the DC bus, no

new electrical or functional failure modes are

created. The temporary battery will be

located in the turbine building, which is nonseismic and a non-vital area. The temporary

battery will not be placed into seismically

mounted racks. Thus, a seismic failure of this

temporary battery is possible. Since the

temporary battery is located in the turbine

building the potential for battery failure to

initiate an accident is not present. The failure

of the temporary battery cannot create a

different response from any previously

postulated accident. Due to the location of

the main turbine-generator in relationship to

the temporary battery, it is not likely that a

turbine missile would strike the battery.

Likewise, an unmitigated Steam Line Break

accident outside the VC would be interrupted

by successful closure of all MSIVs [main

steam isolation valves] thereby leaving the

battery and the associated DC bus intact and

available. This MSIV closure would occur

before any potential steam line break

impacting the battery on the Turbine deck

ensuring necessary DC power to the MSIVs

when needed. Also, any affects of postulated

severe weather on the turbine building have

been evaluated and do not impede the ability

of the remaining DC subsystems to perform

their intended safety function. The remaining

125 VDC EDS and its equipment will

continue to perform the same function and be

operated in the same fashion. The proposed

changes do not introduce any new accident

initiators or precursors, or any new design

assumptions for those systems or

components used to mitigate the

consequences of an accident. Therefore, the

possibility of a new or different kind of

accident from any previously evaluated has

not been created. Thus, the proposed onetime AOT extension TS amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the proposed License Amendment

involve a significant reduction in a margin of

safety?

No. During the replacement of the existing

station batteries, a temporary safety related

battery will perform the same function as the

battery being removed. Even though this

battery is not seismically mounted, not in a

seismically qualified building, nor in a vital

area of the plant it is qualified as a safety

related battery in all other respects.

This battery is virtually identical to the

safety related station battery that is already

installed. It possesses adequate capacity to

fulfill the requirements of the associated 125

VDC bus. The proposed replacement activity

will not prevent the plant from mitigating a

DBA during events that result in the loss of

the temporary battery. In these cases, the

remaining DC power supporting the design

mitigation capability will be maintained. Due

to the limited duration of the activity, the

very low probability of a seismic or other

seismic interaction event over this limited

AOT period and the planned implementing

contingency actions, a significant reduction

in the margin of safety will not result. The

associated DC bus will always be supplied

with both a temporary battery and a battery

charger at all times. The inherent design

conservatism of the 125 VDC system and its

equipment has not been significantly altered;

only the degree of redundancy is not fully

qualified. The 125 VDC EDS and its

equipment will continue to be operated with

the same degree of conservatism.

Accordingly, there is no significant reduction

in the margin of safety.

Therefore, based upon the above

evaluation, the Authority has concluded that

these changes involve no significant hazards

consideration.

The NRC staff has reviewed the licensee’s

analysis and, based on this review, it appears

that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Mr. David E. Blabey,

10 Columbus Circle, New York, New York

10019.

NRC Section Chief: Marsha Gamberoni.

Power Authority of The State of New York,

Docket No. 50–286, Indian Point Nuclear

Generating Unit No. 3, Westchester County,

New York

Date of amendment request: September 7,

2000.

Description of amendment request: The

proposed amendment to the Indian Point

Nuclear Generating Unit No. 3 Technical

Specifications would extend the surveillance

frequency from 720 hours to 1440 hours for

the Fuel Storage Building Emergency

Ventilation 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:

(1) Does the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

evaluated?

Response: The proposed license

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. Extending the surveillance

frequency from 720 hours to 1440 hours

for the Fuel Storage Building Emergency

Ventilation (FSBEV) System charcoal

and HEPA [High Efficiency Particulate

Adsorbers] adsorbers does not involve

any modifications to the plant, will not

require changes to how the plant is

operated nor will it affect the operation

of the plant. Filter systems are not

initiators of accidents, and therefore

extending the filter surveillance

frequency will not increase the

probability of an accident. The way the

filters perform will not be changed by

extending the surveillance frequency. In

addition, it is reasonable to expect

satisfactory filter performance at this

extended frequency based on past

surveillance results. Hence, there is no

change in the assumptions of an

accident. Therefore, this change will not

increase the consequences of an

accident previously evaluated.

(2) Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any accident previously evaluated?

Response: The proposed license

amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. Extending the surveillance

frequency from 720 hours to 1440 hours

for the FSBEV charcoal and HEPA

adsorbers does not involve any

modifications to the plant, will not

require changes to how the plant is

operated nor will it affect the operation

of the plant. Therefore, extending the

surveillance frequency will not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

(3) Does the proposed license

amendment involve a significant

reduction in a margin of safety?

Response: The proposed license

amendment does not involve a

significant reduction in a margin of

safety. Extending the surveillance

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69065

frequency from 720 hours to 1440 hours

for the FSBEV charcoal and HEPA

adsorbers does not change the TS

required methyl iodine efficiency

removal requirement of >90% that

ensures a safety factor of at least 2. This

change is acceptable because it is

reasonable to expect satisfactory filter

performance at this extended frequency

based on past surveillance results,

hence it is reasonable to expect that the

additional 720 hours before testing will

not result in the safety factor being

diminished. Thus, the proposed change

would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Mr. David E.

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Section Chief: Marsha

Gamberoni.

PSEG Nuclear LLC, Docket Nos. 50–272

and 50–311, Salem Nuclear Generating

Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request:

September 26, 2000, as supplemented

on October 6, 2000.

Description of amendment request:

The proposed change would amend the

Salem Nuclear Generating Station

(Salem) Unit Nos. 1 and 2 Technical

Specifications (TSs) to increase the asfound set point tolerance for the

Pressurizer Safety Valves (PSV) from

±1% to ±3%; increase the as-found set

point tolerance for the Main Steam

Safety Valves (MSSV) from ±1% to ±3%;

change the required action for

inoperable MSSVs to require a

reduction in power based upon the

number of inoperable MSSVs, as

opposed to the current requirement to

reduce the Power Range Neutron Flux

High trip setpoint; and remove

specifications and references related to

plant operation with three Reactor

Coolant System loops. The associated

TS Bases sections will also be amended

to reflect the TS 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 any accident previously

evaluated.

Changing the pressurizer and main steam

safety relief valve lift setpoint tolerance from

±1% to ±3% does not significantly increase

the probability of any accident previously

evaluated. The only events initiated by the

opening of these safety valves are the

accidental depressurization of the Reactor

Coolant System and accidental

depressurization of the Main Steam System.

These events are a result of an inadvertent

lifting of these valves and do not depend on

the safety valve lift setpoint or tolerance.

Therefore, the likelihood that either of these

events will occur has not been increased.

[Analyses associated with the limiting

overpressurization transients (Loss of

External Electrical Load and/or Turbine Trip,

and Single Reactor Coolant Pump Locked

Rotor) have been performed that demonstrate

that increasing the Pressurizer Safety Valve

and Main Steam Safety valve lift setpoint

tolerance to ±3% would result in primary

and secondary side pressure responses less

than the acceptance criteria of 110% of the

design pressure. Therefore, since the

proposed setpoint tolerance increase would

not adversely impact current accident

analysis assumptions, the proposed change

would not result in an increase in

consequences of an accident previously

evaluated.]

For operation with inoperable main steam

safety valves, changing the required action

from a reduction of the power range high

neutron flux trip setpoint to a reduction of

the allowable reactor power level will not

increase the consequences of any accident.

With inoperable Main Steam Safety Valves,

the Loss of External Electrical Load and/or

Turbine Trip event becomes limiting in terms

of secondary side pressurization. The high

flux trip does not provide any mitigation for

this event. Other events limiting at power,

that require the power range trip for

mitigation, assume a safety analysis trip

setpoint of 118% (based on a nominal trip

setpoint of 109%) regardless of the initial

power level. Therefore, the proposed change

does not impact any of the accident analysis

assumptions.

The current Salem licensing basis for the

Spurious Activation of the Safety Injection

System credits operator action to unblock a

pressurizer Power Operated Relief Valve

prior to the water solid pressurizer reaching

the safety valve lift setpoint. The analyses

that determined the time at which the safety

valve would reach its pressure setpoint

covered the ¥3% tolerance. Since this would

conservatively result in the earliest opening

time, there was no need to consider the

positive side of the tolerance. The results of

the analyses indicate that the allowable

operator action time is sufficient, such that

water relief occurs through the Power

Operated Relief Valves and not through the

Pressurizer Safety Valves. As such the

consequences of this event have not changed

as a result of the proposed change.

Increasing the Main Steam Safety Valve lift

setting tolerance may result in increased

secondary side backpressure for the

Auxiliary Feedwater Pumps. However,

analyses have demonstrated that with the

elevated backpressures that could result from

increasing the Main Steam Safety Valve

setpoint upper tolerance to +3%, the

Auxiliary Feedwater Pumps would still

provide [greater than the minimum] flow

required to mitigate events in which normal

feedwater is not available, a Loss of Normal

Feedwater and a Loss of Offsite Power to

Station Auxiliaries.

In terms of radiological consequences, the

current design and licensing basis analyses

that include steaming through the Main

Steam Safety Valves bound the proposed lift

setpoint tolerance change.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposal will result in a change in the

allowed Pressurizer Safety Valve and Main

Steam Safety Valve lift setpoint tolerance

range. No physical changes to these valves or

to their nominal lift setpoint is required.

These valves are assumed to malfunction

only as the initiator for the accidental

depressurization of the Reactor Coolant

System or Main Steam System. An increased

lift setpoint tolerance range does not change

the assumption of these depressurization

events nor create a new type of event.

Requiring a reduction in reactor thermal

power in the event of inoperable Main Steam

Safety Valves is consistent with the analysis

methodology. Initiation of any Salem UFSAR

[Updated Final Safety Analysis Report]

analyzed event at a power level less than full

power is bounded by those events analyzed

at full power, or specifically analyzed at the

limiting power level, and does not constitute

a new or different kind of accident. Also, no

changes are being made to the power range

high flux trip setpoint that will make it

inconsistent with any analytical assumption.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

Analyses performed demonstrate that the

proposed increase in the Pressurizer Safety

Valve and Main Steam Safety Valve lift

pressure setpoint tolerance from ±1% to ±3%

will provide acceptable primary and

secondary side pressure responses to the

anticipated operational occurrences and

design basis accidents. The limiting

overpressurization transients, Loss of

External Electrical Load and/or Turbine Trip,

and Single Reactor Coolant Pump Locked

Rotor, stay well within the acceptance

criteria of 110% of the design pressure.

For operation with inoperable Main Steam

Safety Valves, requiring a reduction in

reactor thermal power is consistent with the

accident analysis. The current requirement to

reduce the power range high neutron flux

trip setpoint [does not reduce the] margin of

safety since this trip does not provide any

mitigation for the limiting secondary system

pressurization event, Loss of External

Electrical Load and/or Turbine Trip with

inoperable Main Steam Safety Valves.

69066 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

The current licensing basis for the

Spurious Activation of the Safety Injection

System credits operator action to unblock a

pressurizer Power Operated Relief Valve

prior to the water solid pressurizer reaching

the Pressurizer Safety Valve lift setpoint. As

the Pressurizer Safety Valves are not

designed for water relief, failure to unblock

a Power Operated Relief Valve before

reaching the Pressurizer Safety Valve lift

setpoint would result in water relief and

likely failure of the Pressurizer Safety Valve

to reseat. This condition would escalate the

Spurious Activation of the Safety Injection

System (Condition II event) into a small

break Loss Of Coolant Accident (Condition III

event). The analyses that determined the time

at which primary system pressure would

reach the Pressurizer Safety Valve setpoint

bound the ¥3% tolerance. The results of the

analyses indicate that the allowable operator

action time is sufficient, such that water

relief occurs through the Power Operated

Relief Valves and not through the Pressurizer

Safety Valves. Since the Pressurizer Safety

Valve would not fail due to water relief, there

is no reduction in the margin of safety for

this event.

Increasing the Main Steam Safety Valve lift

setting tolerance may result in increased

secondary side backpressure for the

Auxiliary Feedwater System. However,

analyses have demonstrated that under

degraded Auxiliary Feedwater Pump

performance, and with secondary side

backpressure corresponding to 103% of the

lowest Main Steam Safety Valve setpoint, the

Auxiliary Feedwater System can provide

[greater than the minimum] flow required to

mitigate those events where normal

feedwater is not available, a Loss of Normal

Feedwater and a Loss of Offsite Power to

Station Auxiliaries.

Therefore the proposed changes to the

Technical Specifications 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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Section Chief: James W. Clifford.

Southern California Edison Company,

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

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of amendment requests: October

6, 2000 (PCN–518).

Description of amendment requests:

The amendment application proposes to

revise the San Onofre Nuclear

Generating Station, Units 2 and 3,

Technical Specification (TS) 3.7.11,

‘‘Control Room Emergency Air Cleanup

System (CREACUS)’’ consistent with

generic industry changes recently

approved by the U.S. Nuclear

Regulatory Commission (NRC)

document Technical Specification Task

Force (TSTF)–287. The proposed

amendments would allow up to 24

hours to restore the Control Room

Pressure Boundary (CRPB) to operable

status when two CREACUS trains are

inoperable due to an inoperable CRPB

in MODE 1, 2, 3, or 4. In addition, a

Limiting Condition for Operation note

would be added to allow intermittent

opening of the CRPB under

administrative controls without entering

the Actions.

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 facility in accordance

with the proposed amendments does

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: No.

The Control Room Area Ventilation System

and Control Room boundary are not assumed

to be an initiator of any analyzed accident;

they are provided to minimize doses to the

control room operators during an accident.

Therefore, these proposed changes have no

impact on the probability of occurrence of

any previously analyzed accident.

The proposed changes also have no impact

on offsite dose consequences. The control

room ventilation system and control room

boundary provide protection for control room

personnel and do not mitigate radiological

effluents released offsite. With the control

room boundary inoperable and not

pressurized, the accident analyses assume

unfiltered air would enter the control room

and operator doses would be significantly

increased. Conservative accident analysis

assumptions do not take credit for available

compensatory measures to mitigate operator

dose. Compensatory measures include the

supply of protective clothing, and self

contained breathing apparatus adequate for at

least nine persons within the control room

envelope.

Additionally, for cases where the control

room boundary is opened under

administrative control, appropriate

administrative measures ensure the boundary

can be rapidly restored. Based on the

compensatory measures available to the

control room operator to minimize dose (to

be consistent with the intent of General

Design Criterion 19), the administrative

controls required to rapidly restore an

opened boundary, and considering the low

probability of an event occurring in this short

time period, the consequences are not

considered to be significantly increased.

Operators maintain the ability to mitigate a

design basis event.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated?

Response: No.

No changes are being made to actual plant

hardware which will result in any new

accident causal mechanisms. Therefore, no

new accident causal mechanisms will be

generated.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in a

margin of safety?

Response: No.

Margin of safety is related to the ability of

the fission product barriers to perform their

design functions during and following

accident conditions. These barriers include

the fuel cladding, the reactor coolant system,

and the containment system. The

performance of these barriers will not be

degraded by the proposed changes. The

Control Room Ventilation System and control

room boundary provide a protected

environment for the control room operators

during analyzed events. The proposed

change would allow the boundary to be

degraded for a limited period of time.

However, administrative controls would be

in place to rapidly restore an opened

boundary and existing compensatory

measures (e.g., protective clothing and self

contained breathing apparatus) would be

implemented to minimize operator dose.

Therefore, it is expected that operators would

maintain the ability to mitigate design basis

events and none of the fission product

barriers would be affected by this change.

Therefore, this 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 requests involve no

significant hazards consideration.

Attorney for licensee: Douglas K.

Porter, Esquire, Southern California

Edison Company, 2244 Walnut Grove

Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260 and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2

and 3, Limestone County, Alabama

Date of amendment request: October

6, 2000.

Description of amendment request:

The proposed amendment would amend

each of the three units’ Technical

Specifications (TS) to adopt Technical

Specifications Task Force (TSTF)

change No. 318, Revision 0 (TSTF–318).

TSTF–318 provides a 7-day action

period and completion time in the event

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69067

of inoperability of one of the two low

pressure coolant injection (LPCI) pumps

in each of the two emergency core

cooling system (ECCS) divisions.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed new Condition of one LPCI

pump in each LPCI injection subsystem being

inoperable is more reliable than the current

Limiting Condition for Operation which

allows 2 LPCI pumps in one ECCS subsystem

to be inoperable for 7 days. Also, the LPCI

mode of the Residual Heat Removal system

is not assumed to be initiator of any analyzed

event. Therefore, the proposed amendment

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve a

physical alteration of the plant, add any new

equipment or require any existing equipment

to be operated in a manner different from the

present design. The proposed change will not

impose any new or eliminate any existing

requirements. Therefore, the proposed

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change will not reduce a

margin of safety because it has no effect on

any safety analyses assumptions. The

proposed new Condition for one LPCI pump

in each LPCI injection subsystem represents

a more reliable configuration than the

existing LCO which allows two LPCI pumps

in one ECCS subsystem to be inoperable for

7 days. For these reasons, the proposed

amendment 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: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET I0H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: June 16,

2000, as supplemented by letter dated

September 27, 2000.

Description of amendment request:

The proposed amendments would

revise Technical Specification (TS) 3.7

and TS Tables 3.7–1, 3.7–2, 3.7–3, and

4.1–1. The proposed changes would: (a)

revise the surveillance frequency for

Reactor Protection System and

Engineered Safety Features Actuation

System analog channels from monthly

to quarterly; (b) decrease the frequency

for most permissives to a refueling

interval; (c) increase the time allowed to

perform maintenance on an inoperable

instrument channel; and (d) revise

associated action statements consistent

with NUREG–1431.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Virginia Electric and Power Company has

reviewed the requirements of 10 CFR 50.92

as they relate to the proposed Reactor

Protection System (RPS) and Engineered

Safety Features Actuation System (ESFAS)

Technical Specification changes for the Surry

Units 1 and 2 and determined that a

significant hazards consideration is not

involved. In support of this conclusion, the

following evaluation is provided.

Criterion 1—Operation of Surry Units 1

and 2 in accordance with the proposed

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The determination that the results of the

proposed changes remain within acceptable

criteria was established in the SER(s) [Safety

Evaluation Report(s)] prepared for WCAP–

10271, WCAP–10271 Supplement 1, WCAP–

10271 Supplement 2, WCAP–10271

Supplement 2, Revision 1 and WCAP–14333

issued by letters dated February 21, 1985,

February 22, 1989, April 30, 1998, and July

15, 1998.

Implementation of the proposed changes is

expected to result in an increase in total RPS

and ESFAS yearly unavailability. The

proposed changes have been shown to result

in a small increase in the core damage

frequency (CDF) due to the combined effects

of increased RPS and ESFAS unavailability

and reduced inadvertent reactor trips.

The values determined by the WOG

[Westinghouse Owners Group] and presented

in the WCAP for the increase in CDF were

verified by Brookhaven National Laboratory

(BNL) as part of an audit and sensitivity

analyses for the NRC Staff. Based on the

small value of the increase compared to the

range of uncertainty in the CDF, the increase

is considered acceptable. The analysis

performed by the WOG and presented in the

WCAP included changes to the surveillance

frequencies for the automatic actuation logic

and actuation relays and the reactor trip and

bypass breakers. The overall increase in the

CDF, including the changes to the

surveillance frequencies for the automatic

actuation logic and actuation relays and the

reactor trip and bypass breakers, was

approximately 6 percent. However, even with

this increase, the overall CDF remains lower

than the NRC safety goal of 10 E–4/reactor

year.

Changes to surveillance test frequencies for

the RPS and ESFAS interlocks do not

represent a significant reduction in testing.

The currently specified test interval for

interlock channels allows the surveillance

requirement to be satisfied by verifying that

the permissive logic is in its required state

using the annunciator status light. The

surveillance as currently required only

verifies the status of the permissive logic and

does not address verification of channel

setpoint or operability. The setpoint

verification and channel operability is

verified after a refueling shutdown. The

definition of the channel check includes

comparison of the channel status with other

channels for the same parameter. The

requirement to routinely verify permissive

status is a different consideration than the

availability of trip or actuation channels

which are required to change state on the

occurrence of an event and for which the

function availability is more dependent on

the surveillance interval. Therefore, the

change in the interlock surveillance

requirement to at least once every 18 months

does not represent a significant change in

channel surveillance and does not involve a

significant increase in unavailability of the

RPS and ESFAS.

For the additional relaxations in WCAP–

14333, the WOG evaluated the impact of the

additional relaxation of allowed outage times

and completion times, and action statements

on core damage frequency. The change in

core damage frequency is 3.1 percent for

those plants with two out of three logic

schemes that have not implemented the

proposed surveillance test interval, allowed

outage times, and completion times

evaluated in WCAP–10271 and its

supplements. This analysis calculates a

significantly lower increase in core damage

frequency than the WCAP–10271 analysis

calculated. This can be attributed to more

realistic maintenance intervals used in the

current analysis and crediting the AMSAC

[ATWS (anticipated transient without scram)

mitigating system actuation circuitry] system

as an alternative method of initiating the

auxiliary feedwater pumps. Therefore, the

overall increase in CDF is estimated to be

3.1% for the proposed changes per the

generic Westinghouse analysis.

The NRC performed an independent

evaluation of the impact on core damage

frequency (CDF) and large early release

fraction (LERF). The results of the staff’s

review indicate that the increase in core

damage frequency is small (approximately

3.2%) and the large early release fraction

would increase by only 4 percent for 2 out

of 3 logic schemes that have not

69068 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

implemented the proposed surveillance test

interval, allowed outage times, and

completion times evaluated in WCAP–10271

and its supplements. Further, the absolute

values for CDF still remain within NRC safety

goals.

Therefore, the proposed changes do not

result in a significant increase in the severity

or consequences of an accident previously

evaluated. Implementation of the proposed

changes affects the probability of failure of

the RPS and ESFAS but does not alter the

manner in which protection is afforded or the

manner in which limiting criteria are

established.

Criterion 2—The proposed license

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed changes do not result in a

change in the manner in which the RPS or

ESFAS provide plant protection. No change

is being made which alters the functioning of

the RPS or ESFAS (other than in a test mode).

Rather the likelihood or probability of the

RPS or ESFAS functioning properly is

affected as described above. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident as defined in the Safety Analysis

Report.

The proposed changes do not involve

hardware changes. Some existing

instrumentation is designed to be tested in

bypass and current Technical Specifications

allow testing in bypass. Testing in bypass is

also recognized by IEEE [Institute of

Electrical and Electronics Engineers]

Standards. Therefore, testing in bypass has

been previously approved and

implementation of the proposed changes for

testing in bypass does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Furthermore, since the other proposed

changes do not alter the physical operation

or functioning of the RPS or ESFAS, the

possibility of a new or different kind of

accident from any previously evaluated has

not been created.

Criterion 3—The proposed license

amendment does not involve a significant

reduction in a margin of safety.

The proposed changes do not alter the

safety limits, limiting safety system setpoints

or limiting conditions for operation. The RPS

and ESFAS analog instrumentation remain

operable to mitigate as assumed in the

accident analysis. The impact of reduced

testing other than as addressed above is to

allow a longer time interval over which

instrument uncertainties (e.g., drift) may act.

Implementation of the proposed changes is

expected to result in an overall improvement

in safety by less frequent testing of the RPS

and ESFAS analog instruments and will

result in less inadvertent reactor trips and

actuation of Engineered Safety Features

components.

This analysis demonstrates that the

proposed amendment to the Surry Units 1

and 2 Technical Specifications does not

involve a significant increase in the

probability or consequences of a previously

evaluated accident, does not create the

possibility of a new or different kind of

accident and does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Donald P. Irwin,

Esq., Hunton and Williams, Riverfront

Plaza, East Tower, 951 E. Byrd Street,

Richmond, Virginia 23219.

NRC Section Chief: Richard L. Emch.

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, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852. Publicly

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).

AmerGen Energy Company, LLC,

Docket No. 50–461, Clinton Power

Station, Unit 1, DeWitt County, Illinois

Date of application for amendment:

August 25, 2000, as supplemented

September 21, October 14, and October

25, 2000.

Brief description of amendment: The

amendment revises the reactor vessel

pressure-temperature limits.

Date of issuance: October 31, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 134.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 19, 2000 (65 FR

56598).

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 October 31,

2000.

No significant hazards consideration

comments received: No.

AmerGen Energy Company, LLC,

Docket No. 50–461, Clinton Power

Station, Unit 1, DeWitt County, Illinois

Date of application for amendment:

July 27, 2000, as supplemented October

5, 2000.

Brief description of amendment: The

amendment revises the Safety Limit

Minimum Critical Power Ratio.

Date of issuance: November 3, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 135.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51348).

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 November 3,

2000.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69069

AmerGen Energy Company, LLC, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

March 7, as supplemented on April 21,

June 14, and September 15, 2000.

Brief description of amendment: The

proposed amendment revised the

Technical Specifications to revise the

surveillance requirements from once per

refueling interval for each excess flow

check valve (EFCV) to testing a

representative sample of EFCVs once

per 24 months.

Date of Issuance: October 25, 2000.

Effective date: October 25, 2000 and

shall be implemented within 30 days of

issuance.

Amendment No.: 216.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51354).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated October 25,

2000.

No significant hazards consideration

comments received: No.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

April 7, 2000, as supplemented June 14

and September 11, 2000.

Brief description of amendment: This

amendment revises Technical

Specification (TS) 3/4.7.6, ‘‘Control

Room Emergency Filtration System,’’ TS

3/4.7.7, ‘‘Reactor Auxiliary Building

Emergency Exhaust System,’’ and TS 3/

4.9.12, ‘‘Fuel Handling Building

Emergency Exhaust System.’’

Specifically, these TS have been revised

to provide an action when the Control

Room Emergency Filtration System or

Reactor Auxiliary Building Emergency

Exhaust System ventilation boundary is

inoperable, and a note that allows an

applicable ventilation boundary to be

open intermittently under

administrative controls. The associated

TS Bases are also being changed in

accordance with the amendment. In

addition, TS 3/4.3.3.1, ‘‘Radiation

Monitoring for Plant Operations,’’ has

been modified to provide consistency

between the applicability of the Control

Room Emergency Filtration System and

the radiation monitors that initiate a

control room isolation signal.

Date of issuance: October 30, 2000.

Effective date: October 30, 2000.

Amendment No.: 102.

Facility Operating License No. NPF–

63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 3, 2000 (65 FR 25762). The

supplemental letters dated June 14 and

September 11, 2000, contained

clarifying information only, did not

expand the application beyond the

scope of the initial notice, and did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

October 30, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

February 21, 2000.

Brief description of amendments: The

amendments revised the condensate

storage tank (CST) low-level setpoint to

prevent entrainment of air in the high

pressure coolant injection (HPCI) pump

suction line when taking suction from

the CST. The amendments also revised

the surveillance requirements for the

CST level instruments.

Date of issuance: October 31, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 120 days from the date of

issuance.

Amendment Nos.: 182 and 177.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15376).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 31,

2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

November 23, 1999, as supplemented by

letter dated September 6, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications 5.5.11—Ventilation Filter

Testing Program, which provides the

test requirements for charcoal filters, to

assure compliance with the

requirements of American Society for

Testing and Materials D3803–1989.

Date of issuance: November 2, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 196/177.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15377).

The supplement dated September 6,

2000, provided clarifying information

that did not change the scope of the

November 23, 1999, application and

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 2,

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:

August 10, 2000.

Brief description of amendment: The

amendment revised the Technical

Specifications to allow an alternate

storage configuration of fuel assemblies

adjacent to the walls within Region I of

the spent fuel pool.

Date of issuance: October 24, 2000.

Effective date: As of the date of

issuance to be implemented within 60

days from the date of issuance.

Amendment No.: 224.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 6, 2000 (65 FR

54086).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 24,

2000.

No significant hazards consideration

comments received: No.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade

County, Florida

Date of application for amendments:

July 7, 2000, as supplemented October

2 and 4, 2000.

Brief description of amendments: The

pressure-temperature limits specified in

Technical Specification (TS) 3.4.9.1 and

Figures 3.4–2, 3.4–3 have been

modified, Figure 3.4–4 deleted, and the

Cold Overpressure Mitigation System

(COMS) requirements have been

changed. The COMS is the

Westinghouse version of the Low

Temperature Overpressure Protection

System.

Date of issuance: October 30, 2000.

69070 Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices

Effective date: October 30, 2000.

Amendment Nos.: 208 and 202.

Facility Operating License Nos. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48751).

The supplemental information provided

on October 2 and 4, 2000, provided

clarifying information only and did not

affect the proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 30,

2000.

No significant hazards consideration

comments received: No.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

September 1, 2000.

Brief description of amendments: The

amendments clarify Technical

Specification 3/4.4.4, ‘‘Pressurizer,’’ to

reflect the current power supply to the

pressurizer heaters and require two

operable trains of pressurizer heaters

during Modes 1, 2, and 3. In addition,

the amendments revise the Bases for

Technical Specification 3/4.4.4 to reflect

these changes and clarify the purpose of

the pressurizer heaters.

Date of issuance: October 20, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 45 days.

Amendment Nos.: 246 and 227.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 20, 2000 (65 FR

56952).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 20,

2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

April 19, 2000, as supplemented on

August 31, 2000.

Description of amendment request:

The amendment implements a

performance-based Containment

Leakage Testing Program in accordance

with 10 CFR Part 50, Appendix J,

Option B as a substitute for the

requirements of 10 CFR Part 50,

Appendix J, Option A. The use of this

option requires the implementation of a

program based on Regulatory Guide

1.163, ‘‘Performance-Based Containment

Leak-Test Program,’’ and modification

of the Technical Specifications to reflect

this program.

Date of issuance: November 2, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 186.

Facility Operating License No. NPF–49:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51359).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 2,

2000.

No significant hazards consideration

comments received: No.

Pacific Gas and Electric Company,

Docket No. 50–275, Diablo Canyon

Nuclear Power Plant, Unit No. 1, San

Luis Obispo County, California

Date of application for amendment:

December 31, 1999, as supplemented by

letters dated January 18, July 7,

September 22, and 29, and October 12,

2000.

Brief description of amendment: The

amendment revises Section 2.C.(1) of

Facility Operating License No. DPR–80

to authorize operation of Unit 1 at

reactor core power levels not in excess

of 3411 megawatts thermal (100 percent

rated power). Unit 2 is already

authorized to operate at that power

level. This amendment also revises

several sections within the Improved TS

to reflect the increase in reactor power

level.

Date of issuance: October 26, 2000.

Effective date: October 26, 2000.

Amendment No.: Unit 1—143.

Facility Operating License No. DPR–80:

The amendments revised the Technical

Specifications and operating license.

Date of initial notice in Federal

Register: April 19, 2000 (65 FR 21037).

The January 18, July 7, September 22,

and 29, and October 12, 2000,

supplemental letters provided

additional clarifying information, did

not expand the scope of the application

as originally noticed, and did not

change the staff’s original proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 26,

2000.

No significant hazards consideration

comments received: No.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

July 27, 2000, as supplemented August

16, 2000, and September 29, 2000.

Brief description of amendment: The

amendment provides for the

applicability of the current safety limit

minimum critical power ratio

(SLMCPR), TS Section 1.1.A, to cycles

beyond Cycle 14. The change also

updates the approved version of the

topical report in TS Section 6.9.A.4.b.1.

Date of issuance: October 30, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 266.

Facility Operating License No. DPR–59:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51362).

The August 16 and September 29,

2000, letters provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 30,

2000.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., Docket No. 50–364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request:

September 8, 2000, as supplemented on

October 2, 2000.

Brief Description of amendment: The

amendment revises surveillance

requirements 3.4.11.1 and 3.4.11.4 to

eliminate the requirement to cycle the

Unit 2 pressurizer power-operated relief

valve block valves during the remainder

of operating cycle 14.

Date of issuance: October 25, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: Unit 2—139.

Facility Operating License No. NPF–8:

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: October 10, 2000 (65 FR

60223).

Public comments requested as to

proposed no significant hazards

consideration: Yes.

The notice provided an opportunity to

submit comments on the Commission’s

Federal Register / Vol. 65, No. 221 / Wednesday, November 15, 2000 / Notices 69071

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by November 9,

2000, but indicated that if the

Commission makes a final no significant

hazards consideration determination,

any such hearing would take place after

issuance of the amendment. The

Commission’s related evaluation of the

amendment, finding of exigent

circumstances, and a final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

October 25, 2000.

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:

November 24, 1999 (TS 99–16).

Brief description of amendments:

These amendments revised the

Technical Specifications (TSs) to update

the industry standard that is used to test

the charcoal adsorber efficiency in

safety-related ventilation systems.

Date of issuance: November 2, 2000.

Effective date: November 2, 2000.

Amendment Nos.: 263 and 254.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revised

the TSs.

Date of initial notice in Federal

Register: January 12, 2000 (65 FR 1929).

The September 21, 2000, supplement

provided clarifying information that did

not change the scope of the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 2,

2000.

No significant hazards consideration

comments received: No.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of application for amendment:

September 14, 2000, as supplemented

on September 22, 2000.

Brief description of amendment: The

amendment revises the Technical

Specifications (TSs) to clarify the valve

isolation signal information in the TS

Table 4.7.2 and make an administrative

change to the Table main steam

isolation valves component

identification.

Date of Issuance: October 31, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 30 days.

Amendment No.: 194.

Facility Operating License No. DPR–28:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 27, 2000 (65 FR

58111).

The September 22, 2000,

supplemental letter was within the

scope of the original application and did

not change the staff’s proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated October 31,

2000.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland, this 8th day

of November 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–29250 Filed 11–14–00; 8:45 am]

BILLING CODE 7590–01–P

PENSION BENEFIT GUARANTY

CORPORATION

Interest Assumption for Determining

Variable-Rate Premium; Interest

Assumptions for Multiemployer Plan

Valuations Following Mass Withdrawal

AGENCY: Pension Benefit Guaranty

Corporation.

ACTION: Notice of interest rates and

assumptions.

SUMMARY: This notice informs the public

of the interest rates and assumptions to

be used under certain Pension Benefit

Guaranty Corporation regulations. These

rates and assumptions are published

elsewhere (or are derivable from rates

published elsewhere), but are collected

and published in this notice for the

convenience of the public. Interest rates

are also published on the PBGC’s web

site (www.pbgc.gov).

DATES: The interest rate for determining

the variable-rate premium under part

4006 applies to premium payment years

beginning in November 2000. The

interest assumptions for performing

multiemployer plan valuations

following mass withdrawal under part

4281 apply to valuation dates occurring

in December 2000.

FOR FURTHER INFORMATION CONTACT:

Harold J. Ashner, Assistant General

Counsel, Office of the General Counsel,

Pension Benefit Guaranty Corporation,

1200 K Street, NW., Washington, DC

20005, 202–326–4024. (For TTY/TDD

users, call the Federal relay service tollfree at 1–800–877–8339 and ask to be

connected to 202–326–4024.)

SUPPLEMENTARY INFORMATION:

Variable-Rate Premiums

Section 4006(a)(3)(E)(iii)(II) of the

Employee Retirement Income Security

Act of 1974 (ERISA) and § 4006.4(b)(1)

of the PBGC’s regulation on Premium

Rates (29 CFR part 4006) prescribe use

of an assumed interest rate in

determining a single-employer plan’s

variable-rate premium. The rate is the

‘‘applicable percentage’’ (currently 85

percent) of the annual yield on 30-year

Treasury securities for the month

preceding the beginning of the plan year

for which premiums are being paid (the

‘‘premium payment year’’). The yield

figure is reported in Federal Reserve

Statistical Releases G.13 and H.15.

The assumed interest rate to be used

in determining variable-rate premiums

for premium payment years beginning

in November 2000 is 4.93 percent (i.e.,

85 percent of the 5.80 percent yield

figure for October 2000).

The following table lists the assumed

interest rates to be used in determining

variable-rate premiums for premium

payment years beginning between

December 1999 and November 2000.

For premium payment years

beginning in:

The assumed

interest rate

is:

December 1999 ...................... 5.23

January 2000 .......................... 5.40

February 2000 ........................ 5.64

March 2000 ............................. 5.30

April 2000 ............................... 5.14

May 2000 ................................ 4.97

June 2000 ............................... 5.23

July 2000 ................................ 5.04

August 2000 ........................... 4.97

September 2000 ..................... 4.86

October 2000 .......................... 4.96

November 2000 ...................... 4.93

Multiemployer Plan Valuations

Following Mass Withdrawal

The PBGC’s regulation on Duties of

Plan Sponsor Following Mass

Withdrawal (29 CFR part 4281)

prescribes the use of interest

assumptions under the PBGC’s

regulation on Allocation of Assets in

Single-employer Plans (29 CFR part

4044). The interest assumptions

applicable to valuation dates in

December 2000 under part 4044 are

contained in an amendment to part 4044

published elsewhere in today’s Federal

Register. Tables showing the

assumptions applicable to prior periods

are codified in appendix B to 29 CFR

part 4044.