Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4341

1997, in Conference Room T–2B3,

11545 Rockville Pike, Rockville,

Maryland, has been rescheduled for

February 6–8, 1997. The meeting will

begin at 8:30 a.m. on Thursday,

February 6, 1997, instead of 1:00 p.m.

on Wednesday, February 5, 1997. The

discussion of the item on ‘‘Design-bases

Verification’’ scheduled for Wednesday,

February 5, 1997, has been postponed to

a future meeting as requested by the

NRC staff. All other items pertaining to

this meeting remain the same as

published in the Federal Register on

Thursday, January 23, 1997 (62 FR

3539).

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting Mr. Sam

Duraiswamy, Chief, Nuclear Reactors

Branch (telephone 301/415–7364),

between 7:30 A.M. and 4:15 P.M. EST.

Dated: January 23, 1997.

Andrew L. Bates,

Advisory Committee Management Officer.

[FR Doc. 97–2165 Filed 1–28–97; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from January 4,

1997, through January 16, 1997. The last

biweekly notice was published on

January 15, 1997 (62 FR 2185).

Notice of Consideration of Issuance of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By February 28, 1997, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

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.
4342 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party. 2

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request:

December 30, 1996

Description of amendment request:

The amendment revises (1) chemistry

data (nickel content) shown on

Technical Specification (TS) Figures

3.4-2 and 3.4-3 for TS 3/4.4.9,

‘‘Pressure/Temperature Limits,’’ and (2)

the associated Bases 3/4.4.9 to reflect

changes to chemistry and material

properties and changes to comply with

recent U.S. Nuclear Regulatory

Commission (NRC) rule changes to 10

CFR 50, Appendix G.

Basis for proposed no significant

hazards 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:

This change does not involve a significant

hazards consideration for the following

reasons:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

There are no physical changes to any plant

equipment created by the proposed changes.

The chemistry and material property changes

do not impact the ability of the reactor vessel

to maintain [its] pressure boundary integrity

as previously evaluated. The decrease in EOL

USE [End-of-Life Upper Shelf Energy] for

weld heat 5P6771 is relatively minor and

remains above the required value that has

been prescribed by the NRC to provide the

necessary level of ductility assumed for

reactor vessel integrity evaluations.

Therefore, the accident initiating and

mitigating aspects of the pressure vessel are

not affected. In addition, neither the

proposed change requiring the ISLH [InService Leak and Hydrostatic] test to be

complete before the core is critical nor the

proposed change allowing fuel in the reactor

vessel during ISLH affects any accident

initiating mechanisms. The proposed change

requiring the ISLH test to be completed

before the core is critical will not increase the

consequences of previously evaluated

accidents because it conservatively assures

the core is subcritical. Although the proposed

change allows fuel in the vessel during ISLH

utilizing the ISLH Pressure-Temperature (PT) limits, the consequences of a pressure

boundary leak have not changed because

ISLH testing is already allowed using the

normal plant P-T limits. In addition, the

ISLH will be required to be completed before

the core is allowed to go critical. The

consequences of a leak with fuel in the vessel

during ISLH are the same using either the

normal P-T limits or the ISLH limits.

Therefore, there would be no increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

There are no physical changes to any plant

equipment or new components created by the

proposed changes. The chemistry and

material property changes do not impact the

pressure boundary integrity of the reactor

vessel. The decrease in EOL USE for weld

heat 5P6771 is relatively minor and remains

above the required value that has been

prescribed by the NRC to provide the

necessary level of ductility assumed for

reactor vessel integrity evaluations.

Therefore, the accident initiating aspects of

the pressure vessel are not affected. In

addition, neither the proposed change

requiring the ISLH test to be complete before

the core is critical nor the proposed change

allowing fuel in the reactor vessel during

ISLH creates any new accident initiating

mechanisms.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4343

The changes in chemical and material

properties do not adversely affect any reactor

vessel integrity evaluations, such as PTS

[Pressurized Thermal Shock] or P-T limits.

The USE for weld heat 5P6771 does decrease

slightly as described in TS Bases Table B 3/

4.4-1. However, the predicted EOL USE

remains above the value prescribed in 10

CFR 50, Appendix G and is not a significant

reduction in the margin of safety. With regard

to the proposed changes allowing fuel in the

reactor vessel during ISLH, the existing TS

Bases specifically state that fuel is not to be

in the reactor vessel when the ISLH P-T

curve is utilized. However, this change is

consistent with the revised 10 CFR 50,

Appendix G rule and as such, is not a

significant reduction in the margin of safety.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart,

Acting

Commonwealth Edison Company,

Docket No. 50-010, Dresden Nuclear

Generating Station, Unit 1, Grundy

County, Illinois

Date of amendment request: October

23, 1996

Description of amendment request:

The proposed change would amend the

Dresden Unit 1 Appendix A Technical

Specifications (TS). The proposed

amendment is a complete revision of the

TS to the same format as Dresden Unit

2/3 TS.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. Will operation of the facility according

to this proposed change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. In general the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis (Decommissioning

Plan). Implementation of these changes will

not reduce reliability of equipment assumed

to operate in the current safety analysis

(Decommissioning Plan), or will provide

continued assurance that specified

parameters remain within their acceptance

limits, and as such, will not significantly

increase the probability or consequences of a

previously evaluated accident.

Some of the proposed changes represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. The proposed amendment

for Dresden Station Unit 1’s Technical

Specifications in general is based on STS

[Standard Technical Specifications]

guidelines or NRC accepted changes to other

facilities such as Trojan or San Onofre Unit

1. Any deviations from STS requirements do

not significantly increase the probability or

consequences of any previously evaluated

accidents for Dresden Station Unit 1. The

proposed amendment is consistent with the

current safety analysis (Decommissioning

Plan) and has been previously determined to

represent sufficient requirements for the

assurance and reliability of equipment

assumed to operate in the safety analysis

(Decommissioning Plan), or provide

continued assurance that specified

parameters remain within their acceptance

limits. As such, these changes will not

significantly increase the probability or

consequences of a previously evaluated

accident.

2. Will operation of the facility according

to this proposed change create the possibility

of a new or different kind of accident from

any accident previously evaluated.

No. In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis (Decommissioning

Plan). Others represent minor curtailments of

the current requirements which are based on

generic guidance or previously approved

provisions for other stations. These changes

do not involve revisions to the design of the

station. Some of the changes may involve

revision in the operation of the station;

however, these provide additional

restrictions which are in accordance with the

current safety analysis (Decommissioning

Plan).

The proposed amendment for Dresden

Station Unit 1’s Technical Specifications in

general is based on STS guidelines or NRC

accepted changes to other facilities such as

Trojan or San Onofre Unit 1. The proposed

amendment has been reviewed for

acceptability at the Dresden Nuclear Power

Station considering similarity of system or

component design versus the STS of later

operating plants. Any deviations from STS

requirements do not create the possibility of

a new of different kind of accident previously

evaluated for Dresden Station, Unit 1. No

new modes of operation are introduced by

the proposed changes. The proposed changes

maintain at least the present level of

operability. Therefore, the proposed changes

do not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. Will operation of the facility according

to this proposed change involve a significant

reduction in a margin of safety?

No. In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis (Decommissioning

Plan). Others represent minor curtailments of

the current requirements which are based on

generic guidance or previously approved

provisions for other stations. Some of the

later individual items may introduce minor

reductions in the margin of safety when

compared to the current requirements.

However, other individual changes are the

adoption of new requirements which will

provide significant enhancement of the

reliability of human performance assumed in

the safety analysis (Decommissioning Plan),

or provide enhanced assurance that specified

parameters remain within their acceptance

limits. These enhancements compensate for

the individual minor reductions, such that

taken together, the proposed changes will not

significantly reduce the margin of safety.

The proposed amendment to Technical

Specification Section 6.0 implements present

requirements, or the intent of present

requirements in accordance with the

guidelines set forth in the STS. Any

deviations from STS requirements do not

significantly reduce the margin of safety for

Dresden Station. The proposed changes are

intended to improve readability, usability,

and the understanding of technical

specification requirements while maintaining

acceptable levels of safe operation. The

proposed changes have been evaluated and

found to be acceptable for use at Dresden

based on system design, safety analysis

requirements and operational performance.

Since the proposed changes are based on

NRC accepted provisions at other operating

plants that are applicable at Dresden and

maintain necessary levels of system or

component reliability, the proposed changes

do not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

analysis of the licensee and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450

Attorney for licensee: Michael I.

Miller, Esquire, Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Seymour H.

Weiss

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request:

December 2, 1996

Description of amendment request:

The proposed amendments would
4344 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

revise Technical Specification 3/4.4.2 to

reduce the number of required Safety/

Relief Valves (SRVs). This change will

support a modification to remove five of

the currently installed SRVs due to the

current excess capacity, and to reduce

maintenance costs and worker radiation

dose. The current requirement for 17 of

the 18 installed SRVs to be operable

would be changed to require 12 of the

13 installed SRVs to be operable.

Basis for proposed no significant

hazards 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 because:

The probability of an accident previously

evaluated will not increase as a result of this

change, because the change in valve

configuration, and the accompanying piping

modification does not alter any of the

initiators of an accident or cause them to

occur more frequently. The piping

modifications will be performed consistent

with the current piping classifications for the

affected components. Removal of the SRVs

will not impact the ability of the remaining

SRVs to perform their functions, as described

below.

The consequences of an ASME

Overpressurization Event are not

significantly increased and do not exceed the

previously accepted licensing criteria for this

event. General Electric (GE) has calculated

the revised peak vessel pressure for LaSalle

Station to be 1341 psig, which is below the

1375 psig criterion of the ASME Code for

upset conditions, referenced in Section 5.2.2,

Overpressurization Protection, of the

Updated Final Safety Analysis Report

(UFSAR), and NUREG-0519 (Safety

Evaluation Report related to the operation of

LaSalle County Station, Units 1 and 2, March

1981), and Section 15.2-4, Closure of Main

Steam Isolation Valves (BWR) of NUREG0800 (Standard Review Plan). The

consequences of this event will continue to

be verified on a cycle-specific basis,

beginning with LaSalle Unit 1 Cycle 9

(L1C9). These analysis results will be

approved as part of the normal reload

licensing 10 CFR 50.59 processes.

GE has also performed an analysis of

the limiting Anticipated Transient

Without Scram (ATWS) event, which is

the MSIV Closure Event (MSIVC). This

analysis calculated the peak vessel

pressure to be 1378 psig, which is well

below the 1500 psig criterion of the

ASME Code for emergency conditions.

General Electric has verified that these

results will not be impacted with the

introduction of Siemens fuel.

The conclusions given in the safety

analyses with regards to primary

containment dynamic loads, main steam

piping loads, Loss-of-Coolant Accident

(LOCA) impact, Minimum Critical Power

Ratio (MCPR) impact and SRV availability

also show that current accident and transient

analyses are not impacted by this change

beyond those reanalyzed by GE and

discussed above.

There is no increase in the amount or types

of radioactive release for any of the affected

accidents or transients.

Therefore, there is not a significant

increase in the consequences of an accident

previously evaluated.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The as-left SRV piping configuration will

continue to be consistent with the current

classifications for these piping and supports,

and have been evaluated by Sargent and

Lundy analyses. This ensures no different

types of events may be caused by piping

failures at these locations. This is the only

physical modification proposed by this

submittal, and it will not create the

possibility of a new or different kind of

accident from those previously evaluated.

Other systems are not modified with this

change and have been shown in this

submittal to continue to function as intended

with the new system configuration, with the

exception of the abandoned discharge line

snubbers which may be replaced with struts,

except where they will be retained as

snubbers due to thermal expansion

requirements. The changed supports are

required to function only as struts with the

revised piping. Consideration and evaluation

of this function ensure no new or different

accidents are created.

3) Involve a significant reduction in the

margin of safety because:

While the calculated peak vessel pressures

for the ASME Overpressurization Event and

the MSIVC ATWS Event are increased due to

the proposed SRV removals, the new peak

pressures remain below the respective

licensing acceptance limits associated with

these events.

The actual cycle-specific reload analysis of

the ASME Overpressurization Event will be

verified to be within the licensing acceptance

limit for that event prior to each cycle

startup, as required in the normal reload

10CFR50.59 process. These licensing

acceptance limits have been previously

evaluated as providing a sufficient margin of

safety. For other accidents and transients,

including suppression pool loadings, the

SRV removals have a negligible, if any, effect

on the results, so the margin of safety is

preserved.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

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: August

14, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specification Sections 3.3

(Engineered Safety Features) and 6.9.1.9

(Core Operating Limits Report (COLR));

the basis of Section 3.3, 3.6

(Containment) and 3.10 (Control Rods).

These changes would incorporate the

best estimate approach into the

licensing basis for the Indian Point Unit

No. 2 large break loss-of-coolant

accident (LOCA) analysis.

Basis for proposed no significant

hazards 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:

No physical changes are being made by

this change. The plant conditions assumed in

the analysis are bounded by the design

conditions for all equipment in the plant.

Therefore, there will be no increase in the

probability of a loss-of-coolant accident. The

consequences of a LOCA are not being

increased. That is, it is shown that the

emergency core cooling system is designed so

that its calculated cooling performance

conforms to the criteria contained in 50.46

paragraph b, that is it meets the five criteria

listed in Section II [see application dated

August 14, 1996] of this evaluation. No other

accident is potentially affected by this

change. Therefore, neither the probability nor

the consequences of an accident previously

analyzed is increased due to the proposed

change.

2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any previously

analyzed?

Response:

There are no physical changes being made

to the plant. No new modes of plant

operation are being introduced. The

parameters assumed in the analysis are

within the design limits of existing plant

equipment. All plant systems will perform

equally during the response to a potential

accident. Therefore, the possibility of a new

or different kind of accident than previously

analyzed will not be increased.

3. Does the proposed amendment involve

a significant reduction in the margin of

safety?

Response:

It has been shown that the analytic

technique used in the analysis realistically

describes the expected behavior of the Indian

Point Unit No. 2 reactor system during a
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4345

postulated loss of coolant accident.

Uncertainties have been accounted for as

required by 10 CFR 50.46. A sufficient

number of loss of coolant accidents with

different break sizes, different locations and

other variations in properties have been

analyzed to provide assurance that the most

severe postulated loss of coolant accidents

were calculated. It has been shown by the

analysis that there is a high level of

probability that all criteria contained in 10

CFR 50.46 paragraph b) are met. Therefore

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 50.92(c) are

satisfied.Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Attorney for licensee: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Project Director: S. Singh Bajwa,

Acting Director

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: August

21, 1996

Description of amendment request:

The proposed amendment would

change the licensee’s Technical

Specifications (TSs) Section 3.3.G

(Hydrogen Recombiner System and

Post-Accident Containment Venting

System), the basis for Section 3.3.G, and

Section 4.4, Table 4.4-1 (Containment

Isolation Valves). The change would

remove the existing flame-type

hydrogen recombiners, its supporting

equipment, and replace it with passive

autocatalytic recombiners (PARs). In

addition, the design basis analysis of

post-accident hydrogen generation

would be recalculated.

Basis for proposed no significant

hazards 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?

Neither the probability nor the

consequences of a post-LOCA [loss-of-coolant

accident] combustible gas accident are

increased by the change in recombiners or in

the change to hydrogen generation analysis.

The probability of a 10 CFR 59.44 type LOCA

is not affected. The consequences of such an

accident are not significantly changed.

Accidents associated with failure of the

flame-recombiner flue (hydrogen/oxygen)

system as well as with failure of the flamerecombiner containment isolation valves

have been eliminated.

No other accident is potentially affected by

this change.

2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any previously

evaluated?

No new modes of plant operation are being

introduced other than elimination of

operation of the flame-type recombiners and

associated support equipment. Recombiner

failure is believed to be far less likely with

the PAR design but in any event, the

containment vent system is being maintained

in its current role as backup to recombiner

systems. All other plant systems will perform

equally during the response to a potential

accident. Therefore, the possibility of a new

or different kind of accident than previously

analyzed will not be increased.

3) Does the proposed amendment involve

a significant reduction in the margin of

safety?

The proposed amendment involves margin

in the hydrogen flammability limit, in the

hydrogen generation assumptions and in the

number of PAR devices assumed.

Furthermore, sensitivity analysis on PAR

effectiveness indicates that additional margin

exists for success even with degraded PAR

performance. It has been shown by the

analysis that the criteria of 10 CFR 50.44(d)

can be met with margin. Therefore, 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Attorney for licensee: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Project Director: S. Singh Bajwa,

Acting Director

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: August

22, 1996

Description of amendment request:

The proposed amendment would revise

the licensee’s Technical Specification

Sections 3.3 and 4.5 (Engineered Safety

Features). The proposed revision would

delete the requirement to utilize sodium

hydroxide (NaOH) as an additive in the

posted-accident containment spray

system.

Basis for proposed no significant

hazards 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:

...consistent with the Commission’s criteria

in 10 CFR 50.92, we have determined that the

proposed change does not involve a

significant hazards consideration because the

operation of Indian Point Unit No. 2 in

accordance with this change would not:

1) involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed revisions

are based on conservative analyses utilizing

new, approved methodologies. The analysis

shows the sodium hydroxide spray additive

can be removed without significantly

affecting the radiological consequences of a

postulated LOCA [loss-of-coolant accident]

and that the calculated off-site doses would

remain within the 10 CFR 100 guidelines. In

order to maintain acceptable pH levels in the

recirculating ECC [emergency core cooling]

solution, baskets of trisodium phosphate will

be stored in strategic locations in

containment.

2) create the probability of a new or

different kind of accident from any accident

previously evaluated. The proposed change

allows the containment safeguards to

mitigate the consequences of a design basis

LOCA in a manner equivalent to that

previously approved.

3) involve a significant reduction in a

margin of safety. With the proposed change,

all safety criteria previously evaluated are

still met and remain conservative.

Therefore, based on the above, we

conclude that the proposed changes do not

constitute a significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Attorney for licensee: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Project Director: S. Singh Bajwa,

Acting Director

Duke Power Company, Docket Nos. 50-

413 and 50-414, Catawba Nuclear

Station, Units 1 and 2, York County,

South Carolina

Date of amendment request: January

3, 1997

Description of amendment request:

The proposed amendments would

eliminate from various parts of the

Technical Specifications any

requirement for the low steam pressure
4346 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

signal as an initiator of safety injection.

The licensee stated that the function of

the signal is adequately performed by

other signals (such as the low

pressurizer pressure signal).

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. The proposed change will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change, to delete the SI

[safety injection] signal on low steam line

pressure, will only prevent an unnecessary SI

actuation as an event occurs which involves

secondary system depressurization. No

consequences will significantly increase,

because for each event previously analyzed it

has been shown that either SI on low steam

pressure is not demanded, or that another SI

signal (e.g., low pressurizer pressure) is

generated in sufficient time to meet

applicable acceptance criteria. The

probability of an accident will not increase.

2. The proposed change will not create the

possibility of any new accident not

previously evaluated.

The initiation of SI on a low steam line

pressure signal may occur during events

which involve a depressurization of the

secondary side, including excessive auxiliary

feedwater addition. There are other SI

initiation signals which will accomplish this

same function if needed. Removing this

actuation signal will not create any new

failure modes or necessitate any new

hardware configurations (other than the

deletion of the signal itself). No new accident

scenarios are created.

3. There is no significant reduction in a

margin of safety.

Analysis has shown that for any transient

for which SI would have occurred on low

steam line pressure, transient response is

maintained within acceptable limits. Steam

line break mass and energy releases inside

containment do not violate the existing

environmental qualification envelope. Steam

line breaks outside containment are not

adversely affected by this change.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the proposed

amendments involve no significant

hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

Nos. 1 (ANO-1), Pope County, Arkansas

Date of amendment request:

November 26, 1996

Description of amendment request:

Change Reactor Coolant System

Pressure and Temperature Curvers

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Criterion 1 - Does not Involve a

Significant Increase in the Probability or

Consequences of an Accident Previously

Evaluated.

The proposed change revises the pressure/

temperature limits in accordance with the 10

CFR 50.60 requirements or in accordance

with Code Case N-514. This approach utilizes

the latest NRC guidelines relative to

estimating neutron irradiation damage of the

reactor vessel, as well as maintaining

conservative limits with respect to the low

temperature overpressure protection (LTOP)

system. Therefore, this change does not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed change will not create the

possibility of a new or different kind of

accident from any previously evaluated since

it does not introduce new systems, failure

modes or plant perturbations. Therefore, this

change does not create the possibility of a

new or different kind of accident from any

previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in Margin of Safety.

The proposed change will not involve a

significant reduction in the margin of safety

since the proposed pressure/temperature

limitations have been developed consistent

with the requirements of 10 CFR 50.60. The

operational limits have been developed to

maintain the necessary margins of safety

through 32 effective full power years using

methodologies previously reviewed and

approved by the NRC. The objective of these

limits is to prevent non-ductile failure during

any normal operating condition, including

anticipated operational occurrences and

system hydrostatic tests.

The LTOP safety factors are based on

reanalyzed conditions for 32 effective full

power years of operation utilizing

methodology contained in ASME Code Case

N-514. The LTOP evaluation under Code

Case N-514 for low temperature transients is

considered more appropriate than the ASME

Section XI. The code case establishes a factor

of 110% of the pressure determined to satisfy

Appendix G, paragraph G-2215 of ASME

Section XI, Division 1 as a design limit,

instead of 100% required by Section XI. This

proposed alternative is acceptable because

the Code Case recognizes the conservatism of

the ASME Appendix G curves and allows

establishing a LTOP setpoint which retains

an acceptable margin of safety while

maintaining operational margins for reactor

coolant pump operation at low temperatures

and pressures. The Code Case provides an

acceptable margin of safety against flaw

initiation and reactor vessel failure, and

reduces the potential for an undesired LTOP

actuation. The application of Code Case N514 for ANO-1 will ensure an acceptable

level of safety. Therefore, this change does

not involve a significant reduction in the

margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: October

7, 1996

Description of amendment request:

Modify Plant Protection System Test

Interval to 123 days.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The proposed changes included in this

amendment request are being made to

surveillance intervals, allowances to use

CISAM elements and various administrative

changes. These changes do not alter the

functional characteristics of any plant

component and do not allow any new modes

of operation of any components. These

changes do not involve a significant increase

in the probability of any event initiator to

occur. Therefore, this amendment request

does not involve a significant increase in the

probability of any accident previously

evaluated.

Increasing the surveillance interval for the

RPS and ESFAS instrumentation has two

principal effects with opposing impacts on

risk. The first impact is a slight increase in

core damage frequency that results from the

increased unavailability of the
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4347

instrumentation in question from the

extended testing interval. The unavailability

of the tested instrumentation components is

translated to result in a failure of the reactor

to trip, an anticipated transient without a

scram, or a failure of the appropriate

engineered safety feature to actuate when

required. The opposing impact on risk is the

corresponding reduction in core damage

frequency that would result due to the

reduced exposure of the plant to test induced

transients.

Representative fault tree models were

developed for ANO-2 and the corresponding

core damage frequency increases and

decreases were quantified in CEN-327 and

CEN-327 Supplement 1. The NRC staff found

that changes in the RPS unavailabilities that

result from extending the surveillance test

interval (STI) from 30 days to 90 days were

not considered to be significant. Estimates of

the reduction in scram frequency from the

reduction in test induced scrams and the

corresponding reduction in core damage

frequency were found acceptable. Sequential

testing intervals of 90 days were found to

result in a net reduction in risk.

CE NPSD-576 employed the same

methodology used in CEN-327 and its

supplement to evaluate the impact of

extending the surveillance intervals from

monthly sequential testing to every four

months (triannual) on a staggered test basis.

The corresponding changes in RPS and

ESFAS unavailabilities are quantified in CE

NPSD-576 and are shown to be less than their

counterparts in CEN-327 and its supplement.

Thus, triannual staggered testing should be

acceptable as it results in lower RPS and

ESFAS unavailabilities than for a 90 day test

interval with sequential testing which has

been found to be acceptable to the NRC.

The TS amendment request provided the

option to use cycle independent shape

annealing matrix (CISAM) elements. The

CISAM elements will be validated during

startup testing and will be required to meet

additional acceptance criteria as well as that

used for the cycle specific shape annealing

matrix (SAM) elements. If the CISAM is

determined to be no longer valid, a cycle

specific SAM will be calculated and used in

the CPCs. Therefore, the CPCs will operate as

designed and this change will not affect the

consequences of any accident previously

evaluated.

The CPC addressable constant surveillance

requirements and the various administrative

changes affected by this TS change do not

affect the consequences of any accident

previously evaluated.

Therefore, these changes do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

This amendment request does not involve

any changes in equipment and will not alter

the manner in which the plant will be

operated.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The RPS/ESFAS extended testing interval

yields no significant reduction in the margin

to safety. The instrument drift occurring over

the proposed STI will not cause the setpoint

values to exceed those assumed in the safety

analysis and specified in the TS. There are

no changes to equipment or plant operations

that will result from this change. The

implementation of these proposed changes is

expected to result in an overall improvement

in safety due to the fact that reduced testing

will result in fewer inadvertent trips, less

frequent actuation of EFAS components, and

less frequent distraction of the operations

personnel.

The CPC addressable constant surveillance

interval extension included in this

amendment request is consistent with the

methodology found in NUREG-1432,

‘‘Standard Technical Specifications

Combustion Engineering Plants’’ (ISTS).

Requiring the addressable constant

verification to be performed as part of the

CPC channel functional test should detect an

error in these constants prior to restoring the

channel to operable status instead of

allowing the error to go undetected until the

next surveillance period. Although the

surveillance interval is extended by this TS

change, this change does not involve a

significant reduction in the margin of safety.

The CPC CISAM elements and the various

administrative changes included in this TS

change do not involve a significant reduction

in the margin of safety.

Therefore, these changes do not involve a

significant reduction in the margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request:

December 19, 1996

Description of amendment request:

Change Request Concerning Addition to

the Core Operating Limit Report

References

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Criterion 1 - Does not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The proposed change to add the technical

manual for the Combustion Engineering

Nuclear Transient Simulation (CENTS) code

to the Core Operating Limits Report (COLR)

references is administrative in nature. The

CENTS code has been reviewed and

approved by the NRC. The physical design or

operation of the plant is not impacted by this

proposed change. The proposed change does

not adversely impact transient analysis

assumptions or results. The COLR-related

safety analyses will continue to be performed

utilizing NRC-approved methodologies, and

specific reload changes will be evaluated

under the provisions of 10CFR50.59.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed change to reference the NRCapproved CENTS code is administrative in

nature. No physical alterations of plant

configuration, changes to plant operating

procedures, or operating parameters are

proposed. No new equipment is being

introduced, and no equipment is being

operated in a manner inconsistent with its

design. The COLR-related safety analyses

will continue to be performed utilizing NRCapproved methodologies. A 10CFR50.59

safety review will continue to be performed

to evaluate specific reload changes.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed change to reference the

CENTS code is administrative in nature.

Existing technical specification operability

and surveillance requirements are not

reduced by the proposed change. The cyclespecific COLR limits for future reloads will

continue to be developed based on NRCapproved methodologies. Technical

specifications will continue to require that

the core be operated within these limits and

specify appropriate actions to be taken if the

limits are violated. The COLR-related safety

analyses will continue to be performed

utilizing NRC-approved methodologies, and

specific reload changes will be evaluated per

10CFR50.59. Therefore, this change does not

involve a significant reduction in the margin

of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve a significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff
4348 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request:

December 19, 1996

Description of amendment request:

Change Request Concerning Power

Calibration Requirements

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Criterion 1 - Does not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The proposed change will redefine the

tolerance band allowed for linear power

level, the Core Protection Calculator (CPC)

delta T Power, and CPC nuclear power

signals. Changing the tolerance range from

[plus or minus] 2% to between -0.5% and

10% between 15% and 80% rated thermal

power, will require more conservative

tolerances than are currently allowed. This

change will ensure that the power

indications are more conservative relative to

the existing safety analyses. Therefore, this

change does not involve a significant

increase in the probability or consequences

of any accident previously evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed change to Technical

Specification power calibration tolerance

limits are conservative relative to the current

requirements. This amendment request does

not change the design or operation of any

plant systems or components. Therefore, this

change does not create the possibility of a

new or different kind of accident from any

previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in Margin of Safety.

The allowed tolerance band for the linear

power level, CPC delta T power, and CPC

nuclear power signals between 15 and 80%

power has been redefined. The new

requirements are more conservative than the

tolerances that currently exist in the

Technical Specifications. This change will

ensure that the power indications are more

conservative relative to the existing safety

analyses. Therefore, this change does not

involve a significant reduction in the margin

of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request:

December 19, 1996

Description of amendment request:

Change Request Concerning Reactor

Coolant System Volume

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Criterion 1 - Does not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

This proposed change allows the relocation

of the reactor coolant system volume in the

design features section of technical

specifications to the safety analysis report.

Future changes will be controlled under

10CFR50.59. This change is considered

administrative in nature. Appropriate values

of reactor coolant system volume are used in

the safety analyses. This change does not

affect any system or component functional

requirements. The operation of the plant is

not affected by this change.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The relocation of existing requirements

from the technical specifications to another

licensee controlled document is

administrative in nature. This change does

not modify or remove any plant design

requirement. The proposed change will not

affect any plant system or structure, nor will

it affect any system functional or operability

requirements. Therefore, no new failure

modes are introduced as a result of this

change.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed amendment request relocates

the coolant system volume located in the

technical specifications design feature

section to another licensee controlled

document, the ANO-2 Safety Analysis

Report, which is controlled under

10CFR50.59. The proposed change is

administrative in nature because the design

requirements for the facility remain the same.

The proposed change does not represent a

change in the configuration or operation of

the plant.

Therefore, this change does not involve a

significant reduction in the margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve a significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request:

December 19, 1996

Description of amendment request:

Change Control Room Ventillation

System Requirements

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Criterion 1 - Does not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The control room emergency ventilation

and air conditioning systems are not

initiators of an accident previously

evaluated. Extension of the allowable outage

time for one inoperable control room

emergency air conditioning system from 7

days to 30 days is acceptable based on the

low probability of an event occurring that

would require control room isolation and a

concurrent or subsequent failure of the

remaining operable control room emergency

air conditioning system. An evaluation using

probabilistic safety assessment techniques

has shown the frequency of this event to be

an acceptably low level (4.67E-6/yr). The

ANO-1 surveillance requirements for the

control room emergency ventilation and air
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4349

conditioning system have been updated for

consistency with the ANO-2 requirements

and are consistent with RG 1.52, March 1978,

Revision 2 and ASTM D3803-1989. The

change in the ANO-2 Mode of Applicability

for the control room radiation monitoring

instrumentation is acceptable because the

only identified accident scenario requiring

control room isolation on high radiation

while in Modes 5 and 6 is the fuel handling

accident and this analysis shows that the

dose consequences to the control room

operators are acceptable in the event of a fuel

handling accident, assuming that the normal

control room ventilation system is properly

isolated. The remainder of the changes have

been made for consistency between the ANO1 and ANO-2 TS and are considered to be

more restrictive or administrative in nature.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The control room emergency ventilation

and air conditioning systems are not accident

initiators. The proposed changes introduce

no new mode of plant operation and no new

possibility for an accident is introduced by

modifying the ANO-1 surveillance testing

requirements for the control room emergency

ventilation and air conditioning systems.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

With the exception of the AOT extension

and the relaxation of the ANO-2 Mode of

Applicability for the control room radiation

monitoring instrumentation, all the ANO-1

and ANO-2 changes are considered

administrative or more restrictive and are

intended to clarify and make consistent the

requirements of the control room emergency

habitability equipment. Although the AOT

extension does involve an incremental

reduction in the margin of safety due to slight

increase in the frequency of an event

requiring control room isolation, followed by

failure of the operable emergency control

room chiller, a probabilistic safety

assessment has shown this slight increase in

frequency (approximately 3.58E-6/yr) to be

acceptably low. The change in the ANO-2

Mode of Applicability for the control room

radiation monitoring instrumentation is

acceptable because the only identified

accident scenario requiring control room

isolation on high radiation while in Modes 5

and 6 is the fuel handling accident and this

analysis shows that the dose consequences to

the control room operators are acceptable in

the event of a fuel handling accident,

assuming that the normal control room

ventilation system is properly isolated.

Therefore, this change does not involve a

significant reduction in the margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket No. 50-366, Edwin I.

Hatch Nuclear Plant, Unit 2, Appling

County, Georgia

Date of amendment request:

December 3, 1996

Description of amendment request:

The two proposed changes would revise

Technical Specification (TS) 2.1.1.2 for

Hatch Nuclear Plant, Unit 2, Safety

Limit Minimum Critical Power Ratio

(SLMCPR) values. The revision is based

upon unique plant evaluations for the

current Cycle 13 and the use of General

Electric (GE) GE-13 fuel, a 9 x 9 fuel

design, in the next Cycle 14. The

proposed SLMCPRs for Hatch Unit 2 are

1.08 and 1.09 (single-loop operation) for

the current Cycle 13, and 1.12 and 1.14

(single-loop operation) for Cycle 14.

The new SLMCPRs were calculated

using NRC-approved methods and

interim implementing procedures. The

SLMCPRs are set high enough to ensure

that greater than 99.9% of all fuel rods

in the core avoid transition boiling if the

limit is not violated. The SLMCPRs

incorporate a margin for uncertainty in

the core operating state for uncertainties

that are fuel-type dependent, including

fuel bundle nuclear characteristics,

critical power correlation, and

manufacturing tolerances. These interim

procedures were revised to incorporate

the following cycle-specific parameters:

(1) Actual core loading, (2) Conservative

variations of projected control blade

patterns, (3) Actual bundle parameters

(e.g., local peaking), and (4) Full cycle

exposure range.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration which

is presented below:

1. Does the change involve a significant

increase in the probability of consequences of

an accident previously evaluated?

The derivation of the revised SLMCPRs for

Plant Hatch Unit 2 for incorporation into the

Technical Specifications, and its use to

determine cycle-specific thermal limits, were

performed using NRC-approved methods.

Additionally, interim implementing

procedures incorporating cycle-specific

parameters were used. Based upon the use of

these calculations, revised SLMCPRs cannot

increase the probability or severity of an

accident. The basis of the SLMCPR

calculation is to ensure that ≤ 99.9% of all

fuel rods in the core avoid transition boiling

if the limit is not violated. The new

SLMCPRs preserve the existing margin to

transition boiling and fuel damage in the

event of a postulated accident. Thus, it can

be concluded that the probability of fuel

damage is not increased and the proposed

Technical Specifications changes do not

involve an increase in the probability or

consequences of an accident evaluation.

2. Do the proposed changes create the

possibility of a new or different type of

accident from any previously evaluated?

The SLMCPR is a Technical Specifications

numerical value designed to ensure that fuel

damage from transition boiling does not

occur as a result of the limiting postulated

accident. The SLMCPRs were calculated

using NRC-approved methods. Additionally,

interim procedures incorporating cyclespecific parameters were used in the

analysis. Therefore, the proposed changes do

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Do the proposed changes involve a

significant reduction in the margin of safety?

The margin of safety as defined in the

Bases will remain the same. The new

SLMCPRs were calculated using NRCapproved methods which are in accordance

with the current fuel design and licensing

criteria. Additionally, interim implementing

procedures, which incorporate cycle-specific

parameters were used. The SLMCPR remains

high enough to ensure that ≤ 99.9% of all fuel

rods in the core will avoid transition boiling

if the limit is not violated, thereby preserving

the fuel cladding integrity. 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.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Herbert N.

Berkow
4350 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of amendment request: January

7, 1997

Description of amendment request:

The proposed amendments would

change the Technical Specifications

(TS) for Plant Hatch, Units 1 and 2,

associated with Surveillance

Requirement (SR) testing that requires

manually actuating every safety/relief

valve (S/RV) during each unit startup

from a refueling outage. The proposed

changes would provide an alternate

method of testing the S/RVs during

shutdown conditions rather than during

unit startup as is currently done. This

approach would reduce valve leakage,

thereby reducing the possibility of

inadvertent valve actuation and

resultant plant transients. Additionally,

deletion of testing for the safety mode of

the S/RVs is proposed since other

testing provides operability verification.

Furthermore, the licensee proposes

relief from the applicable requirements

of the ASME OM Code (1995),

Appendix I, paragraph I 3.4.1(d), which

also requires manual actuating of S/RVs

during unit startup.

Current Unit 1 and Unit 2 SRs

3.5.1.12 and 3.6.1.6.1 require that each

S/RV be manually actuated at pressure

conditions. Georgia Power Company

(GPC) proposes to revise SRs 3.5.1.12

and 3.6.1.6.1 that would require the S/

RVs to be manually actuated in the

relief mode during a plant outage before

steam is generated. The solenoid valve

would be energized, the actuator would

stroke, and the pilot rod lift would be

measured. This in-situ test would verify

that, given a signal to the solenoid, the

pilot disc rod would lift. If steam were

present, the pilot disc would open and

initiate opening of the main stage.

The licensee also proposes to delete

current Units 1 and 2 SR 3.4.3.2, which

also requires that each S/RV be

manually actuated because this test is

not necessary to assure S/RV operability

in the safety mode since other tests,

taken together, confirm the entire S/RV

assembly functions adequately.

Basis for proposed no significant

hazards 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:

Georgia Power Company [GPC] has

reviewed the proposed license amendment

request and determined its adoption does not

involve a significant hazards consideration.

In support of this determination, an

evaluation of each of the three 10 CFR 50.92

standards follows.

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Since the proposed Technical

Specifications changes and ASME Code relief

do not impose any physical changes to the

S/RVs, their design function is unaffected.

The submittal only proposes changes to the

manner in which the S/RVs are tested. As

discussed in Enclosure 1 [of the licensee’s

submittal], the combination of current S/RV

testing and the proposed alternate

testing will continue to adequately

demonstrate the operability of the S/RVs for

both the safety and relief modes. Under the

proposed testing requirements, it is expected

that S/RV leakage will decrease; thus, the

probability of occurrence of an inadvertent S/

RV actuation is actually reduced.

FSAR [Final Safety Analysis Report]

analyzed events, such as MSIV [main steam

isolation valve] closure, generator load reject,

turbine trip with failure of switchyard

breakers to open, and pressure regulator

failure, take credit for the S/RVs mitigating

the consequences of these events. These

proposed changes will not increase the

consequences of these events, since a series

of S/RV tests (on the bench and installed)

will ensure all S/RV components necessary

to ensure valve opening will function. The S/

RVs will therefore be capable of performing

their design functions.

Furthermore, reducing the number of

manual actuations of the S/RVs decreases the

likelihood of a stuck open S/RV, which is an

analyzed event in the Hatch FSAR.

Therefore, the probability of occurrence

and the consequences of previously analyzed

events are not increased.

2. The proposed changes do not create the

possibility of [a new or different kind of

accident from any accident] previously

evaluated.

The proposed changes affect the manner in

which S/RV operability is verified in that one

Technical Specifications SR [surveillance

requirement] is being deleted and two are

being revised; however, they do not affect the

way the S/RVs are operated. The S/RVs will

not be operated or tested in a manner

contrary to their design. As a result, no new

mode of operation is introduced. That is, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

The present method of S/RV testing

unnecessarily challenges the valves, and is

linked to S/RV degradation through pilot

valve and/or main valve leakage. This

Technical Specifications change should

decrease S/RV leakage and improve S/RV

reliability by reducing the potential for

spurious valve actuation at full power. In this

sense, the margin of safety is actually

increased; e.g., the likelihood for spurious S/

RV actuation is reduced.

Deleting the test of installed S/RVs at rated

temperature and pressure will not

significantly reduce the margin of safety for

events in which S/RV actuation is assumed,

since each S/RV receives a series of tests

which insure each component necessary for

successful opening of the S/RV functions

properly. Thus, the S/RV is assured of

opening in either the safety or the relief

mode. For example, at Wyle Labs, the valves

undergo testing at operating steam pressure.

This test ensures operability of the pilot and

main discs and also verifies set pressure,

reseat pressure, and main steam stroke time.

As noted previously, upon successful

completion of these tests, including

verification of zero seat leakage, the valves

receive a written certification from the lab

and are returned to Plant Hatch for

installation.

GPC further proposes that, upon

installation, but before steam is generated,

the valves receive a test requiring the

solenoid to be energized. This test provides

additional verification that the pilot disc

opens. The remaining segments of the S/RV

tests verify the ability of ADS and LLS logic

to energize the solenoid.

In summary, this amendment does not

involve a significant reduction in the margin

of safety, because of the reduction in S/RV

degradation, and because remaining tests

confirm the valves will function properly

when required.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Herbert N.

Berkow

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request:

December 16, 1996

Description of amendment request:

The amendment request, if approved,

would reflect the change in the legal

name of the operator of TMI-1 from GPU

Nuclear Corporation to GPU Nuclear

Inc. and reflect in the TMI-1 license and

the Technical Specifications the

registered trade name of GPU Energy.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration (SHC),

which is presented below:

GPU Nuclear Inc. has determined that the

proposed TMI-1 license amendment and

technical specification change request
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4351

involve no significant hazards consideration

as defined in 10 CFR 50.92 because:

Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The proposed amendment adds to

the license and the technical specifications

the trade name of the Owners of TMI-1. The

change in the legal name of the operator of

TMI-1 is a cosmetic change made to reflect

the name changes made throughout the GPU

family of companies. The name change has

no impact on plant design or operation.

Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because no new failure

modes are created by the proposed changes.

The use of a common trade name for the

Owners of TMI-1 and the change in the legal

name of the operator of TMI-1 has no impact

on plant design or operation. Thus, there is

no creation of the possibility of a new or

different kind of accident from those

previously evaluated.

Operation of the facility in accordance

with the proposed amendment will not

involve a signficant reduction in a margin of

safety. The proposed amendment does not

change any operating limits for reactor

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.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037. NRC Acting

Project Director: Patrick D. Milano

Indiana Michigan Power Company,

Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien

County, Michigan

Date of amendment request: August 4,

1995 as supplemented December 20,

1996 [AEP:NRC:1129E and 1129M]

Description of amendment request:

The proposed amendment would

modify the technical specifications (T/

S) to allow for repair of hybrid

expansion joint (HEJ) sleeves under

redefined repair boundary limits. This

alternate plugging criterion would

assess the integrity of parent tube

indications based on the degraded joint

geometry, with reference to the specific

location of the flaw. The continued

operability of the HEJ sleeved tube

would be based on the measured

diameter difference, or diameter delta

(delta D), between the sleeve peak

hardroll diameter and the diameter of

the sleeve adjacent to the parent tube

flaw in the upper joint.

Basis for proposed no significant

hazards 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:

Conformance of the proposed amendments

to the standards for a determination of no

significant hazard as defined in 10 CFR 50.92

(three factor test) is shown in the following.

(1) Operation of Cook Nuclear Plant unit 1

in accordance with the proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The HEJ sleeved tube structural integrity

limits defined by this amendment provide for

structural integrity consistent with the

guidance of RG 1.121. Tube structural

integrity consistent with the most limiting

RG 1.121 loading is inherently provided by

a measured [delta D] of less than 1 mil,

although the criterion specifies a minimum

of 3 mils must be verified. The structural

integrity characteristics of a postulated

degraded parent tube with a 3 mil [delta D]

provides for axial restraint capability of more

than double the most limiting RG 1.121

loading, which indicates that the postulated

separated tube would not become axially

displaced relative to the sleeve during any

plant condition.

Based on tube pull data from Cook Nuclear

Plant and other plants it is expected that TSP

intersections would provide a substantial

axial restraint capability. This interaction is

neglected in the analysis of the criterion, and

provides for extra safety margin.

Based on the destructive examination

results for sections of HEJ sleeved tubes

removed in 1994 from another plant, the

parent tube flaw morphology is described as

circumferentially oriented with multiple

initiation sites. This segmented morphology

indicates that the previously performed

structural capability testing is conservative.

Additional axial load bearing capability is

provided by the segmented morphology since

end cap loading would be transmitted

through the tube by the non-degraded

ligaments of the segmented crack network,

and tube separation therefore, is not likely or

credible.

The consequences of any postulated failure

of a sleeved tube to which the criteria has

been applied would be bounded by the

current steam generator tube rupture event

discussed in the Cook Nuclear Plant Final

Safety Analysis Report (FSAR). Axial

displacement of any tube, sleeved or

unsleeved, is bounded by approximately 1.1

inch. A tube which experiences axial

displacement by this amount would be

expected to exhibit a release rate well below

the normal makeup capacity. In order for a

HEJ sleeved tube to exhibit reactor coolant

system release rates approaching the release

rates assumed in the FSAR the tube must be

displaced by approximately 3 inches. In

order for the postulated separated tube to

experience axial displacement of any

magnitude, it must be assumed that the HEJ

hardroll provides no structural benefit and

that the tube-to-TSP interaction is

frictionless.

Postulated primary to secondary leakage

during a main steam line break event will be

assessed against the limit of 8.4 gpm in the

faulted loop, calculated as part of the voltage

based plugging limit for tube support plate

intersections. The total of all leakage sources

must be shown to be less than this value.

Application of the 3 mil [delta D] criterion

(excluding eddy current uncertainty) does

not change existing reactor coolant system

flow conditions, therefore, existing LOCA

analysis results will be unaffected. Plant

response to design basis accidents for the

current tube plugging and flow conditions

are not affected by the repair process; no new

tube diameter restriction is introduced.

(2) The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Application of the proposed 3 mil [delta D]

HEJ sleeved tube structural integrity criterion

will not introduce significant or adverse

changes to the plant design basis. The 3 mil

[delta D] criteria provides for structural

integrity of the HEJ sleeved tube assembly

which significantly exceeds the limiting RG

1.121 loading condition. Under these

conditions neither a single nor a multiple

tube rupture event is considered credible.

The general outline of the HEJ sleeve is

unaffected, and the application of the

proposed criterion does not change the sleeve

configuration or size/shape. The application

of the criterion also does not represent a

potential to affect other plant components.

(3) The proposed license amendment does

not involve a significant reduction in a

margin of safety.

The proposed criterion has been shown to

provide structural integrity of the tube

bundle consistent with the most limiting RG

1.121 tube integrity recommendations. In

order for tube rupture to occur, the degraded

parent tube must experience a complete

circumferential separation and be

subsequently axially displaced by

approximately 3 inches. The inherent

structural integrity provided by the

interference fit of the HEJ in addition to the

axial restraint provided by tube support plate

intersections above the HEJ provides for

structural integrity far exceeding the RG

1.121 loading of 2264 lb. Even in the event

that a degraded HEJ sleeved parent tube were

to experience axial displacement, the

maximum amount of displacement the tube

could experience is bounded by 1.11 inch.

Postulating that the tube were to become

displaced by this amount, primary to

secondary leakage would be limited to well

less than the normal makeup capacity due to

the proximity between the hydraulically

expanded sleeve OD and tube ID.

Pulled HEJ sleeved tube samples from

another plant with HEJ sleeved tubes indicate

that the crack morphology is described as

circumferentially oriented cracking with

multiple initiation sites. This segmented
4352 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

morphology provides for additional

structural margin not modeled in the testing

program.

Existing flow equivalency calculations for

the HEJ sleeved tubes will be unaffected by

the application of the criterion.

Based on the preceding analysis it is

concluded that operation of Cook Nuclear

Plant unit 1 following the application of the

3 mil [delta D] HEJ sleeved tube structural

integrity limit does not increase the

probability of an accident previously

evaluated, create the possibility of a new or

different kind of accident from any accident

previously evaluated, or reduce any margins

to plant safety. Therefore, the license

amendment does not involve a significant

hazards consideration as defined in 10 CFR

50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

No. 50-278, Peach Bottom Atomic

Power Station, Unit No. 3, York County,

Pennsylvania

Date of application for amendment:

October 30, 1996

Description of amendment request:

These amendments revise the safety

limit minimum critical power ratios

(SLMCPRs) at Peach Bottom Atomic

Power Station, Unit 3.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1) The proposed TS [technical

specification] changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The derivation of the cycle-specific

SLMCPRs for incorporation into the TS, and

its use to determine cycle-specific thermal

limits, have been performed using USNRC

[U.S. Nuclear Regulatory Commission]-

approved methods as discussed in ‘‘General

Electric Standard Application for Reactor

Fuel,’’ NEDE-24011-P-A-11, and U.S.

Supplement, NEDE-24011-P-A-11-US,

November 17, 1995 and interim

(reconfirmation) implementing procedures.

This change in SLMCPRs cannot increase the

probability or severity of an accident.

The basis of the SLMCPR calculation is to

ensure that greater than 99.9% of all fuel rods

in the core avoid transition boiling if the

limit is not violated. The new SLMCPRs

preserve the existing margin to transition

boiling and fuel damage in the event of a

postulated accident. The fuel licensing

acceptance criteria for the SLMCPR

calculation apply to PBAPS [Peach Bottom

Atomic Power Station], Unit 3, Cycle 11 in

the same manner as they have applied

previously. The probability of fuel damage is

not increased. Therefore, the proposed TS

changes do not involve an increase in the

probability or consequences of an accident

previously evaluated.

2) The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The SLMCPR is a TS numerical value,

designed to ensure that transition boiling

does not occur in 99.9% of all fuel rods in

the core during the limiting postulated

accident. It cannot create the possibility of

any new type of accident. The new SLMCPRs

are calculated using USNRC-approved

methods (≥General Electric Standard

Application for Reactor Fuel,’’ NEDE-24011-

P-A-11, and U.S. Supplement, NEDE-24011-

P-A-11-US, November 17, 1995) and interim

(reconfirmation) implementing procedures.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident, from any accident

previously evaluated.

3) The proposed TS changes do not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the TS

Bases will remain the same. The new

SLMCPRs are calculated using USNRCapproved methods (≥General Electric

Standard Application for Reactor Fuel,’’

NEDE-24011-P-A-11, and U.S. Supplement,

NEDE-24011-P-A-11-US, November 17, 1995)

and interim (reconfirmation) implementing

procedures which are in accordance with the

current fuel licensing criteria. The SLMCPRs

ensure that greater than 99.9% of all fuel rods

in the core will avoid transition boiling if the

limit is not violated, thereby preserving the

fuel cladding integrity. Therefore, the

proposed TS changes do not involve a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Power Authority of the State of New

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

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: January

15, 1997

Description of amendment request:

The amendment proposes to relocate the

snubber operability, surveillance, and

record requirements for components

(snubbers) in the Technical

Specifications (TS) to plant controlled

documents.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

based on the following:

1. These changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

The changes relocate operability,

surveillance, and record requirements for

components (snubbers) which do not meet

the criteria for inclusion in the Technical

Specifications (TS). The affected components

are not assumed to be initiators of analyzed

events and are not assumed to mitigate

accident or transient events. The snubber

requirements will be relocated from the TS

to plant controlled documents. These

requirements will be maintained pursuant to

10 CFR 50.59. Therefore, the changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The changes do not create the possibility

of a new or different type of accident

previously evaluated because:

The changes do not necessitate a physical

alteration of the plant (no new or different

type of equipment will be installed) or affect

parameters governing normal plant

operation. Adequate control of future

changes to snubber requirements will be

maintained. Thus, these changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated for the plant.

3. The proposed changes do not involve a

reduction in a margin of safety because:

The changes do not involve a change to the

operability, surveillance, and record

requirements for the snubber program as they

currently exist in the TS, nor do they impact

on any safety analysis assumptions. The

proposed changes relocate snubber

requirements from the TS to plant controlled

documents. Changes to the requirements in

these documents are subject to the

requirements of 10 CFR 50.59. In addition,

exceptions to code requirements for testing

will require NRC approval. Regulations and
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4353

FitzPatrick commitments to the NRC contain

the necessary programmatic requirements for

the plant controlled documents. Operating

limitations will continue to be imposed, and

required surveillances will continue to be

performed in accordance with regulations,

FitzPatrick commitments to the NRC, and

written procedures and instructions that are

auditable by the NRC. If snubber

inoperability causes a TS system or

component to be inoperable, then the

affected system or component Limiting

Condition for Operation (LCO) will be

entered. Based on the above, the proposed

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Acting Director

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: January

7, 1997

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.2.5 to

incorporate an exception to the

provisions of TS 4.0.4 and to clarify the

time at which the surveillance can be

performed by adding that the

surveillance is to be performed within

24 hours after attaining steady state

conditions at or above 90% rated

thermal power. The revised surveillance

would also contain editorial

enhancements that do not change the

intent of the current surveillance. TS

Table 3.2-1 for Salem Unit 1 would be

revised to delete reference to three loop

operation (which is not permitted at

Salem Unit 1) in order to eliminate

potential confusion when applying this

table.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The changes proposed on the RCS [Reactor

Coolant System] flow measurement and

exemption to Specification 4.0.4 do not affect

the operation of the equipment during

conditions when they are required to perform

their safety function. No physical changes to

the plant result from the proposed changes

made to the surveillance requirements. The

measurement of RCS flow does not impact

the probability of an accident.

Testing is being performed with the plant

in the condition in which the automatic

initiation signals for low RCS flow would

result in a time consistent with the TS

requirements.

Protection System in providing a reactor

trip upon a loss of RCS flow. Degradations in

flow will occur over a long duration;

however, testing will continue to be

performed within twenty-four hours upon

achieving steady state greater than or equal

to 90% RTP [Rated Thermal Power] after

refueling which is a sufficiently short

duration after startup to identify flow

degradations.

Changes proposed to refer to Table 3.2-1

for the DNB [Departure from Nucleate

Boiling] parameters and to delete the Unit 1

three loop operation parameters, and the

inclusion of the type of test performed are

editorial in nature.

Therefore, the consequences of an accident

previously evaluated are not significantly

increased by the proposed changes.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not involve any

modifications to existing plant equipment, do

not alter the function of any plant systems,

do not introduce any new operating

configurations or new modes of plant

operation, nor change the safety analyses.

The point at which RCS flow is measured

using a heat balance will not impact the

ability to maintain or monitor Reactor

Coolant flows. The proposed changes will,

therefore, not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

Changes proposed to refer to Table 3.2-1

for the DNB parameters and to delete the

Unit 1 three loop operation parameters, and

the inclusion of the type of test performed are

editorial in nature.

[The proposed changes will, therefore, 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 changes to the RCS flow surveillance

do not decrease the scope of the existing

testing, but will clarify the point at which the

testing is performed.

The time in which testing is performed,

after achieving steady state conditions after

reaching greater than or equal to 90% RTP

ensures that testing is performed in a timely

manner. Flow margins established as a result

of previous testing will not be significantly

reduced in light of recent outage activities.

Future changes that might impact margins

established by the testing will be reviewed in

accordance with the requirements of 10 CFR

50.59.

Changes proposed to refer to Table 3.2-1

for the DNB parameters and to delete the

Unit 1 three loop operation parameters, and

the inclusion of the type of test performed are

editorial in nature.

All changes are consistent with the intent

of Salem’s current TS [Technical

Specification] and with the 18 month

surveillances specified in NUREG-1431,

Revision 1.

The proposed change, therefore, does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

Southern Nuclear Operating Company,

Inc., Docket No. 50-348, Joseph M.

Farley Nuclear Plant, Unit 1, Houston

County, Alabama

Date of amendments request:

December 26, 1996

Description of amendments request:

The proposed amendment would revise

Technical Specification 3/4.4.6 ‘‘Steam

Generators’’ and its associated Bases.

Specifically, the steam generator repair

limit would be modified to clarify that

the appropriate method for determining

serviceability for tubes with outside

diameter stress corrosion cracking at the

tube support plate is by a methodology

that more reliably assesses structural

integrity. This amendment request is in

accordance with NRC’s Generic Letter

95-05, ‘‘Voltage-Based Repair Criteria

for Westinghouse Steam Generator

Tubes Affected by Outside Diameter

Stress Corrosion Cracking.’’

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1) Operation of Farley units in accordance

with the proposed license amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Testing of model boiler specimens for free

standing tubes at room temperature

conditions shows burst pressures as high as

approximately 5000 psi for indications of

outer diameter stress corrosion cracking with

voltage measurements as high as 26.5 volts.

Burst testing performed on pulled tubes,
4354 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

including tubes pulled from Farley Unit 1,

with up to 7.5 volt indications show burst

pressures in excess of 5300 psi at room

temperature. ... [T]ube burst criteria are

inherently satisfied during normal operating

conditions by the presence of the tube

support plate. Furthermore, correcting for the

effects of temperature on material properties

and minimum strength levels (as the burst

testing was done at room temperature), tube

burst capability significantly exceeds the R.G.

[Regulatory Guide] 1.121 criterion requiring

the maintenance of a margin of 1.43 times the

steam line break pressure differential on tube

burst if through-wall cracks are present

without regard to the presence of the tube

support plate. Considering the existing data

base, this criterion is satisfied with bobbin

coil indications with signal amplitudes over

twice the 2.0 volt voltage-based repair

criteria, regardless of the indicated depth

measurement. This structural limit is based

on a lower 95% confidence level limit of the

data at operating temperatures. The 2.0 volt

criterion provides an extremely conservative

margin of safety to the structural limit

considering expected growth rates of outside

diameter stress corrosion cracking at Farley.

Alternate crack morphologies can correspond

to a voltage so that a unique crack length is

not defined by a burst pressure to voltage

correlation. However, relative to expected

leakage during normal operating conditions,

no field leakage has been reported from tubes

with indications with a voltage level of under

7.7 volts for a 3/4 inch tube with a 10 volt

correlation to 7/8 inch tubing (as compared

to the 2.0 volt proposed voltage-based tube

repair limit). Thus, the proposed amendment

does not involve a significant increase in the

probability or consequences of an accident.

Relative to the expected leakage during

accident condition loadings, the accidents

that are affected by primary-to-secondary

leakage and steam release to the environment

are Loss of External Electrical Load and/or

Turbine Trip, Loss of All AC Power to

Station Auxiliaries, Major Secondary System

Pipe Failure, Steam Generator Tube Rupture,

Reactor Coolant Pump Locked Rotor, and

Rupture of a Control Rod Drive Mechanism

Housing. Of these, the Major Secondary

System Pipe Failure is the most limiting for

Farley in considering the potential for off-site

doses. The offsite dose analyses for the other

events which model primary-to-secondary

leakage and steam releases from the

secondary side to the environment assume

that the secondary side remains intact. The

steam generator tubes are not subjected to a

sustained increase in differential pressure, as

is the case following a steam line break event.

This increase in differential pressure is

responsible for the postulated increase in

leakage and associated offsite doses following

a steam line break event. In addition, the

steam line break event results in a bypass of

containment for steam generator leakage.

Upon implementation of the voltage-based

repair criteria, it must be verified that the

expected distributions of cracking

indications at the tube support plate

intersections are such that primary-tosecondary leakage would result in site

boundary dose within the current licensing

basis. Data indicate that a threshold voltage

of 2.8 volts could result in through-wall

cracks long enough to leak at steam line

break conditions. Application of the

proposed repair criteria requires that the

current distribution of a number of

indications versus voltage be obtained during

the refueling outages. The current voltage is

then combined with the rate of change in

voltage measurement and a voltage

measurement uncertainty to establish an end

of cycle voltage distribution and, thus, leak

rate during steam line break pressure

differential. The leak rate during a steam line

break is further increased by a factor related

to the probability of detection of the flaws.

If it is found that the potential steam line

break leakage for degraded intersections

planned to be left in service coupled with the

reduced allowable specific activity levels

result in radiological consequences outside

the current licensing basis, then additional

tubes will be plugged or repaired to reduce

steam line break leakage potential to within

the acceptance limit. Thus, the consequences

of the most limiting design basis accident are

constrained to present licensing basis limits,

and therefore there is no change to the

probability or consequences of an accident

previously evaluated.

2) The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Implementation of the proposed voltagebased tube repair criteria does not introduce

any significant changes to the plant design

basis. Use of the criteria does not provide a

mechanism that could result in an accident

outside of the region of the tube support plate

elevations. Neither a single or multiple tube

rupture event would be expected in a steam

generator in which the repair criteria have

been applied during all plant conditions. The

bobbin probe signal amplitude repair criteria

are established such that operational leakage

or excessive leakage during a postulated

steam line break condition is not anticipated.

Southern Nuclear has previously

implemented a maximum leakage limit of

140 gpd per steam generator. The R.G. 1.121

criterion for establishing operational leakage

limits that require plant shutdown are based

upon leak-before-break considerations to

detect a free span crack before potential tube

rupture. The 140 gpd limit provides for

leakage detection and plant shutdown in the

event of the occurrence of an unexpected

single crack resulting in leakage that is

associated with the longest permissible crack

length. R.G. 1.121 acceptance criteria for

establishing operating leakage limits are

based on leak-before-break considerations

such that plant shutdown is initiated if the

leakage associated with the longest

permissible crack is exceeded. The longest

permissible crack is the length that provides

a factor of safety of 1.43 against bursting at

steam line break pressure differential. A

voltage amplitude of approximately 9 volts

for typical outside diameter stress corrosion

cracking corresponds to meeting this tube

burst requirement at the 95% prediction

interval on the burst correlation. Alternate

crack morphologies can correspond to a

voltage so that a unique crack length is not

defined by the burst pressure versus voltage

correlation. Consequently, a typical burst

pressure versus through-wall crack length

correlation is used below to define the

‘‘longest permissible crack’’ for evaluating

operating leakage limits.

The single through-wall crack lengths that

result in tube burst at 1.43 times steam line

break pressure differential and steam line

break conditions are about 0.54 inch and 0.84

inch, respectively. Normal leakage for these

crack lengths would range from about 0.4

gallons per minute to 4.5 gallons per minute,

respectively, while lower 95% confidence

level leak rates would range from about 0.06

gallons per minute to 0.6 gallons per minute,

respectively.

An operating leak rate of 140 gpd per steam

generator has been implemented. This

leakage limit provides for detection of 0.4

inch long cracks at nominal leak rates and 0.6

inch long cracks at the lower 95% confidence

level leak rates. Thus, the 140 gpd limit

provides for plant shutdown prior to

reaching critical crack lengths for steam line

break conditions at leak rates less than a

lower 95% confidence level and for three

times normal operating pressure differential

at less than nominal leak rates.

Considering the above, the implementation

of voltage-based repair criteria will not create

the possibility of a new or different kind of

accident from any previously evaluated.

3) The proposed license amendment does

not involve a significant reduction in margin

of safety.

The use of the voltage-based repair criteria

is demonstrated to maintain steam generator

tube integrity commensurate with the

requirements of Generic Letter 95-05 and

R.G. 1.121. R.G. 1.121 describes a method

acceptable to the NRC staff for meeting GDC

[General Design Criteria] 2, 14, 15, 31, and 32

by reducing the probability of the

consequences of steam generator tube

rupture. This is accomplished by

determining the limiting conditions of

degradation of steam generator tubing, as

established by inservice inspection, for

which tubes with unacceptable cracking

should be removed from service. Upon

implementation of the criteria, even under

the worst case conditions, the occurrence of

outside diameter stress corrosion cracking at

the tube support plate elevations is not

expected to lead to a steam generator tube

rupture event during normal or faulted plant

conditions. The most limiting effect would be

a possible increase in leakage during a steam

line break event. Excessive leakage during a

steam line break event, however, is

precluded by verifying that, once the criteria

are applied, the expected end of cycle

distribution of crack indications at the tube

support plate elevations would result in

minimal, and acceptable primary to

secondary leakage during the event and,

hence, help to demonstrate radiological

conditions are less than an appropriate

fraction of the 10 CFR [Part] 100 guideline.

The margin to burst for the tubes using the

voltage-based repair criteria is comparable to

that currently provided by existing technical

specifications.

In addressing the combined effects of

LOCA [loss-of-coolant accident] + SSE [safe

shutdown earthquake] on the steam generator
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4355

component (as required by GDC 2), it has

been determined that tube collapse may

occur in the steam generators at some plants.

This is the case as the tube support plates

may become deformed as a result of lateral

loads at the wedge supports at the periphery

of the plate due to either the LOCA

rarefaction wave and/or SSE loadings. Then,

the resulting pressure differential on the

deformed tubes may cause some of the tubes

to collapse.

There are two issues associated with steam

generator tube collapse. First, the collapse of

steam generator tubing reduces the RCS

[reactor coolant system] flow area through

the tubes. The reduction in flow area

increases the resistance to flow of steam from

the core during a LOCA which, in turn, may

potentially increase Peak Clad Temperature

(PCT). Second, there is a potential the partial

through-wall cracks in tubes could progress

to through-wall cracks during tube

deformation or collapse or that short throughwall indications would leak at significantly

higher leak rates than included in the leak

rate assessments.

Consequently, a detailed leak-before-break

analysis was performed and it was concluded

that the leak-before-break methodology (as

permitted by GDC 4) is applicable to the

Farley reactor coolant system primary loops

and, thus, the probability of breaks in the

primary loop piping is sufficiently low that

they need not be considered in the structural

design basis of the plant. Excluding breaks in

the RCS primary loops, the LOCA loads from

the large branch line breaks were analyzed at

Farley and were found to be of insufficient

magnitude to result in steam generator tube

collapse or significant deformation.

Regardless of whether or not leak-beforebreak is applied to the primary loop piping

at Farley, any flow area reduction is expected

to be minimal (much less than 1%) and PCT

margin is available to account for this

potential effect. Based on analyses’ results,

no tubes near wedge locations are expected

to collapse or deform to the degree that

secondary to primary in-leakage would be

increased over current expected levels. For

all other steam generator tubes, the

possibility of secondary-to-primary leakage

in the event of a LOCA + SSE event is not

significant. In actuality, the amount of

secondary-to-primary leakage in the event of

a LOCA + SSE is expected to be less than that

originally allowed, i.e., 500 gpd per steam

generator. Furthermore, secondary-toprimary in-leakage would be less than

primary-to-secondary leakage for the same

pressure differential since the cracks would

tend to tighten under a secondary-to-primary

pressure differential. Also, the presence of

the tube support plate is expected to reduce

the amount of in-leakage.

Addressing the R.G. 1.83 considerations,

implementation of the tube repair criteria is

supplemented by 100% inspection

requirements at the tube support plate

elevations having outside diameter stress

corrosion cracking indications, reduced

operating leakage limits, eddy current

inspection guidelines to provide consistency

in voltage normalization, and rotating probe

inspection requirements for the larger

indications left in service to characterize the

principle degradation mechanism as outside

diameter stress corrosion cracking.

As noted previously, implementation of

the voltage-based repair criteria will decrease

the number of tubes that must be taken out

of service with tube plugs or repaired. The

installation of steam generator tube plugs or

tube sleeves would reduce the RCS flow

margin, thus implementation of the voltagebased repair criteria will maintain the margin

of flow that would otherwise be reduced

through increased tube plugging or sleeving.

Considering the above, it is concluded that

the proposed change does not result in a

significant reduction in margin with respect

to plant safety as defined in the Final Safety

Analysis Report or any bases of the plant

Technical Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow

Southern Nuclear Operating Company,

Inc., Docket No. 50-348, Joseph M.

Farley Nuclear Plant, Unit 1, Houston

County, Alabama

Date of amendments request: January

10, 1997

Description of amendments request:

The proposed amendments would

implement repair of tubes using laser

welded tube sleeves for the steam

generators at Farley Units 1 and 2 as

described in WCAP-13088, Revision 4,

and WCAP-14740. In addition, for Unit

2, references to a one-cycle limited

implementation of L* are being

removed. The approval for the limited

implementation of L* expired at the last

Unit 2 outage in the fall of 1996.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. Operation of Farley Units 1 and 2 in

accordance with the proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The laser welded sleeve configurations as

described within WCAP-13088, Revision 4

and WCAP-14740 have been designed and

analyzed in accordance with the

requirements of the ASME Code [American

Society of Mechanical Engineers Boiler and

Pressure Vessel Code]. Fatigue and stress

analyses of the sleeved tube assemblies

produced acceptable results. Mechanical

testing has shown that the structural strength

of Alloy 690 sleeves under normal, faulted

and upset conditions is within acceptable

limits. Leakage testing for 7/8 inch tube

sleeves has demonstrated that significant

primary-to-secondary leakage is notexpected

during all plant conditions, including the

case where the seal weld is not produced in

the lower joint of the tubesheet sleeve.

Initial acceptance of welded joints uses

ultrasonic inspection to verify that all weld

thicknesses meet the minimum specified

conditions over the entire circumference. A

plugging limit of 24% allowable depth of

penetration of the sleeve tube wall thickness

applies for each type of laser welded sleeve

that may be installed in the Farley Nuclear

Plant steam generators and is determined for

uprated conditions with a limiting steam

pressure for reduced Thot and 20% steam

generator tube plugging conditions. These

conditions represent the limiting primary-tosecondary operating pressure differential,

which is bounding for the sleeve plugging

limit and structural analysis inputs.

However, the state-of-the-art in eddy current

inspection capability is such that no probes

are qualified to size the depth of penetration

of stress corrosion cracking. It is generally

believed that the detection threshold of these

probes is well below 40% throughwall.

Southern Nuclear Operating Company will

plug on detection any crack-like indications

that may occur in the sleeve using the sleeve

inspection probe of record until an

inspection process is qualified to size depth

of penetration of stress corrosion cracking

into the tube wall.

The hypothetical consequences of failure

of the sleeve would be bounded by the

current steam generator tube rupture analysis

included in the Farley Nuclear Plant FSAR

[Final Safety Analysis Report]. Due to the

slight reduction in diameter caused by the

sleeve wall thickness, it is expected that

primary coolant release rates would be

slightly less than assumed for the steam

generator tube rupture analysis (depending

on the break location), and therefore, would

result in lower total primary fluid mass

release to the secondary system.

Combinations of tubesheet sleeves and tube

support plate sleeves would reduce the

primary fluid flow through the sleeved tube

assembly due to the series of diameter

reductions the fluid would have to pass on

its way to the break area. The overall effect

would be reduced steam generator tube

rupture release rates.

As addressed previously, the proposed

Technical Specification change to support

the installation of full length tubesheet,

elevated tubesheet, or tube support plate

elevation Alloy 690 laser welded sleeves as

described in WCAP-13088, Revision 4 and

WCAP-14740 does not adversely impact any

other previously evaluated design basis

accident or the results of LOCA [lossof-coolant accident] and non-LOCA

accident analyses for the current Technical

Specification minimum reactor coolant
4356 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

system flow rate. The results of the analyses

and testing, as well as plant operating

experience, demonstrate that the sleeve

assembly is an acceptable means of restoring

tube integrity to a condition consistent with

its original design basis. Also, per Regulatory

Guide 1.83, Revision 1 recommendations, the

condition of the sleeved tube can be

monitored through periodic inspections with

present eddy current techniques.

Conformance of the sleeve design with the

applicable sections of the ASME Code and

results of the leakage and mechanical tests

support the conclusion that the installation

of laser welded tube sleeves will not increase

the probability or consequences of an

accident previously evaluated. Depending

upon the break location for a postulated

steam generator tube rupture event,

implementation of tube sleeving could act to

reduce the radiological consequences to the

public due to reduced primary to secondary

flow rate through a sleeved tube compared to

a non-sleeved tube based on the restriction

afforded by the sleeve wall thickness.

Removal of the references to the interim

use of an L* repair criteria will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Implementation of laser welded sleeving

will not introduce significant or adverse

changes to the plant design basis. Sleeving

also does not represent a potential to affect

any other plant component. Stress and

fatigue analysis of the repair has shown the

ASME Code minimum stress values are not

exceeded. Implementation of laser welded

sleeving maintains overall tube bundle

structural and leakage integrity at a level

consistent to that of the originally supplied

tubing during all plant conditions. Leak and

mechanical testing of sleeves support the

conclusions of the calculations that each

sleeve joint retains both structural and

leakage integrity during all conditions.

Sleeving of tubes does not provide a

mechanism resulting in an accident outside

of the area affected by the sleeves. Any

hypothetical accident as a result of potential

tube or sleeve degradation in the repaired

portion of the tube is bounded by the existing

tube rupture accident analysis. Since the

sleeve design does not affect any other

component or location of the tube outside of

the immediate area repaired, in addition to

the fact that the installation of sleeves and

the impact on current plugging level analyses

is accounted for, the possibility that laser

welded sleeving creates a new or different

type of accident is not credible.

Removal of the references to the interim

use of an L* repair criteria will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed license amendment does

not involve a significant reduction in margin

of safety.

The laser welded sleeving repair of

degraded steam generator tubes as identified

in WCAP-13088, Revision 4, has been shown

by analysis to restore the integrity of the tube

bundle consistent with its original design

basis condition as the requirements of the

ASME Code are satisfied. The safety factors

used in the design of sleeves for the repair

of degraded tubes are consistent with the

safety factors in the ASME Boiler and

Pressure Vessel Code used in steam generator

design. The design of the tubesheet sleeve

lower joints for the 7/8 inch sleeves (for both

the full length and elevated tubesheet sleeve)

have been verified by testing to preclude

realistic leakage during normal and

postulated accident conditions.

The portions of the installed sleeve

assembly which represent the reactor coolant

pressure boundary can be monitored for the

initiation and progression of sleeve/tube wall

degradation, thus satisfying the

recommendations of Regulatory Guide 1.83,

Revision 1 and the surveillance requirements

included in Specification 4.4.6.0. The portion

of the tube bridged by the sleeve joints is

effectively removed from the pressure

boundary, and the sleeve then forms the new

pressure boundary. The areas of the sleeved

tube assembly which require inspection are

defined in WCAP-13088, Revision 4.

The effect of sleeving on the design

transients and accident analyses have been

reviewed based on the installation of sleeves

up to the level of steam generator tube

plugging coincident with the minimum

reactor flow rate. The installation of sleeves

is to be evaluated as the equivalent of some

level of steam generator tube plugging.

Evaluation of the installation of sleeves is

based on the determination that LOCA

evaluations for the licensed minimum reactor

coolant flow bound the effect of a

combination of tube plugging and sleeving

up to an equivalent of the actual steam

generator tube plugging limit. Information

provided in WCAP-13088, Revision 4,

describes the method to determine the flow

equivalency for all combinations of tubesheet

and tube support plate sleeves in order that

the minimum flow requirements are met.

Implementation of laser welded sleeving

will reduce the potential for primary-tosecondary leakage during a postulated steam

line break while maintaining available

primary coolant flow area in the event of a

LOCA. By effectively isolating degraded areas

of the tube through repair, primary pressure

boundary integrity is restored and the

potential for primary-to-secondary leakage

during all plant conditions is minimized.

These degraded tubes are returned to a

condition consistent with the design basis.

While the installation of a sleeve causes a

reduction in primary coolant flow, the

reduction is significantly below the reduction

incurred by plugging. Therefore, greater

primary coolant flow area is maintained

through sleeving.

Removal of the references to the interim

use of an L* repair criteria will not involve

a significant reduction in margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow

Tennessee Valley Authority, Docket No.

50-390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: January

10, 1997

Description of amendment request:

The proposed amendment would

modify the Watts Bar Nuclear Plant

(WBN) Unit 1 Technical Specifications

(TS) in order to implement the 1995 rule

change to 10 CFR Part 50, Appendix J.

The revised Appendix J provided an

Option B which allows performance

based testing for containment leakage

rate testing. The TS in Section 3.6 and

associated Bases, TS Section 3.0.2 and

TS Section 5.7 would be changed. Also,

the schedular exemption for

containment airlock testing now

specified in the facility license in

Section 2.D(1) would no longer be

required and would be deleted.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration,

which is presented below:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment to WBN TSs is

in accordance with Option B to 10 CFR 50,

Appendix J. The proposed amendment adds

a voluntary performance-based option for

containment leak-rate testing. The changes

being proposed do not affect the precursor for

an accident or transient analyzed in Chapter

15 of WBN Final Safety Analysis Report. The

proposed change does not increase the total

allowable primary containment leakage rate.

The proposed change does not reflect a

revision to the physical design and/or

operation of the plant. [T]herefore, operation

of the facility, in accordance with the

proposed change, does not significantly affect

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment to WBN TSs is

in accordance with the new performancebased option (Option B) to 10 CFR 50,
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4357

Appendix J. The changes being proposed will

not change the physical plant or the modes

of operation defined in the facility license.

The proposed changes do not increase the

total allowable primary containment leakage

rate. The changes do not involve the addition

or modification of equipment, nor do they

alter the design or operation of plant systems.

Therefore, operation of the facility in

accordance with the propsoed change does

not create the possibility of a new or diferent

kind of accident from any previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in margin of

safety.

The proposed change to WBN TSs is in

accordance with the new option to 10 CFR

50, Appendix J. The proposed option is

formulated to adopt performance-based

approaches. This option removes the current

prescriptive details from the TS. The

proposed changes do not affect plant safety

analyses or change the physical design or

operation of the plant. The proposed change

does not increase the total allowable primary

containment leakage rate. Therefore,

operation of the facility, in accordance with

the proposed change, does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request:

November 26, 1996, as supplemented

December 17, 1996

Description of amendment request:

The proposed amendments would allow

a one-time only change necessary to

replace the existing 125-volt dc battery

cells with new cells. Date of publication

of individual notice in Federal Register:

December 13, 1996 (61 FR 65605)

Expiration date of individual notice:

January 13, 1997

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

Notice Of Issuance Of Amendments To

Facility Operating LIcenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

August 23, 1996

Brief description of amendment: This

amendment makes Technical

Specifications changes allowing fuel

enrichment of up to 5.0 weight percent

Uranium-235. The previous limit was

4.1 weight percent. This change allows

Arkansas Nuclear One, Unit-2, to

receive, store, and use nuclear fuel of

5.0 weight percent Uraninum-235.

Date of issuance: January 14, 1997

Effective date: January 14, 1997

Amendment No.: 178

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 9, 1996 (61 FR 52964)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated January 14, 1997.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

November 24, 1996, as supplemented on

December 2, 1996.

Brief description of amendment: This

amendment adds small break-loss-of

coolant accident methodology CENPD137, Supplement 1-P and its approval

letter by the NRC as a reference to

Section 6.9.5.1. This code previously

approved by the NRC increases the

steam generator tube plugging limit to

30% with an associated reduction of

10% in RCS flow. This amendment also

corrects a typographical error in

Specification 6.9.5.1.8, and

Specifications 6.9.5.1.10 through

6.9.5.1.14 are numbered to

accommodate these changes.

Date of issuance: January 14, 1997

Effective date: January 14, 1997

Amendment No.: 179

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 3, 1996 (61 FR

64173) However, on December 9, 1996,
4358 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

the licensee verified that the number of

plugged tubes would not exceed their

current 10% limit established by the old

code. This determination removed the

basis for considering this request as

exigent. Since the potential does exist

for the plugging to exceed the 10% in

the future, the technical specification

amendment request is therefore, a valid

request on a normal schedule. This

change did not alter the staff’s initial

proposed no safety hazard condition

determination, therefore noticing was

not warranted. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

January 14, 1997.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

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

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: October

24, 1996

Brief description of amendment: The

amendment revises the technical

specifications to delete the accelerated

testing requirements for the standby

diesel generators. This action is

consistent with the provisions of

Generic Letter 94-01, ‘‘Removal of

Accelerated Testing and Special

Reporting Requiremets for Emergency

Diesel Generators,’’ dated May 31, 1994.

Date of issuance: January 14, 1997

Effective date: January 14, 1997

Amendment No.: 90

Facility Operating License No. NPF47. The amendment revised the

Technical Specifications/operating

license.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64384) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 14, 1997.No significant hazards

consideration comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

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

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: May 30,

1996

Brief description of amendment: The

amendment revises the technical

specification surveillance requirement

3.8.3.4 to specify a 5-start pressure for

the air recievers associated with the

Division III, High Pressure Core Spray

emergency diesel generator.

Date of issuance: January 16, 1997

Effective date: January 16, 1997

Amendment No.: 91

Facility Operating License No. NPF47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34892) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated January 16, 1997.No

significant hazards consideration

comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

September 23, 1996

Brief description of amendment:

Changes to Technical Specification (TS)

to delete a note for the Surveillance

Requirement 3.3.7.1 for the Engineered

Safeguard Actuation System Logic.Date

of issuance: January 6, 1997

Effective date: January 6, 1997

Amendment No.: 155

Facility Operating License No. DPR72. Amendment revised the TS.

Date of initial notice in Federal

Register: October 23, 1966 (61 FR

55034) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 6, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: October

23, 1996, as supplemented by letter

dated November 6, 1996.

Brief description of amendments: The

amendments revised Technical

Specification 3.4.6.1, regarding reactor

coolant system leakage detection

instrumentation, to adopt the

requirements found in NUREG-1431,

‘‘Standard Technical Specifications

Westinghouse Plants,’’ for the reactor

coolant system leakage detection

instrumentation.

Date of issuance: January 8, 1997

Effective date: January 8, 1997

Amendment Nos.: Unit 1 -

Amendment No. 86; Unit 2 -

Amendment No. 73

Facility Operating License Nos. NPF76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64387) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 8, 1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

April 19, 1996, and as supplemented on

August 15, 1996

Brief description of amendment: The

amendment introduces new Technical

Specification (TS) 3.10.10, ‘‘Single

Control Rod Withdrawal - Refueling,’’

under TS 3.10, ‘‘SPECIAL

OPERATIONS.’’ The purpose of this

Special Operations LCO is to permit the

withdrawal of a single control rod for

testing in MODE 5 without imposing the

requirements for establishing the

secondary containment and main

control room boundaries as normally

required during Core Alterations.

Date of issuance: January 13, 1997

Effective date: January 13, 1997

Amendment No.: 112

Facility Operating License No. NPF62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25707)

and September 25, 1996 (61 FR 50344).

The August 15, 1996, submittal changed

the focus of the original amendment

request, therefore, it was re-noticed in

the FEDERAL REGISTER. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated January 13, 1997.No

significant hazards consideration

comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

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:

September 12, 1995
Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices 4359

Brief description of amendment: The

amendment revises Technical

Specification 6.3.1 to add a requirement

that the Assistant Operations Manager

hold a senior reactor operator (SRO)

license if the Operations Manager does

not hold an SRO license for Millstone

Unit 3.

Date of issuance: January 7, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 132

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13530)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated January 7, 1997.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

April 26, 1996, as supplemented August

23, 1996.

Brief description of amendment: The

amendment changes requirements

regarding reactor coolant system leakage

testing following refueling outage and

other sytem pressure testing of reactor

coolant system following repairs,

replacements, or modifications.

Date of issuance: January 7, 1997

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 171

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28602)

The August 23, 1996, letter provided

clarifying information that did not

change the initial no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated January 7, 1997No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Rochester Gas and Electric

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

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

October 29, 1996

Brief description of amendment: The

amendment corrects an error with

respect to Table 3.3.2-1, Function 6c of

the Technical Specifications (TSs)

which references the incorrect Required

Action for inoperable channels of the

auxiliary feedwater pump actuation on

Steam Generator Level - Low Low logic.

The TSs are revised to correct the

Required Action to place the inoperable

channel in ‘‘trip’’ within 6 hours or

initiate a plant shutdown to Mode 4.

Date of issuance: January 9, 1997

Effective date: January 9, 1997

Amendment No.: 66

Facility Operating License No. DPR18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64395) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 9, 1997No significant hazards

consideration comments received: No

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

Rochester Gas and Electric

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

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

October 29, 1996

Brief description of amendment: This

amendment revises the MODE of

applicability for the motor-driven

auxiliary feedwater pump actuation on

opening of the main feedwater pump

breakers to correct an error introduced

during Amendment No. 61.

Date of issuance: January 9, 1997

Effective date: January 9, 1997

Amendment No.: 67

Facility Operating License No. DPR18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64395) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 9, 1997No significant hazards

consideration comments received: No

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

Wisconsin Electric Power Company,

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

Beach Nuclear Plant, Unit Nos. 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

April 29, 1996, as supplemented

October 21, December 2, and December

16, 1996

Brief description of amendments:

These amendments revise Technical

Specification (TS) Section 15.3.14, ‘‘Fire

Protection System,’’ and Section

15.4.15, ‘‘Fire Protection System,’’ and

relocate the requirements of the fire

protection program from the TS and

incorporate, by reference, the NRCapproved fire protection program into

the Final Safety Analysis Report. In

addition, the amendments revise the

operating licenses to include the NRC’s

standard fire protection condition. The

amendments also approve

administrative changes consistent with

the relocation as well as corrections to

several typographical errors.

Date of issuance: January 8, 1997

Effective date: January 8, 1997, and

implementation within 90 days from the

date of issuance. Implementation shall

include the relocation of Technical

Specification requirements to the

appropriate licensee-controlled

document as identified in the licensee’s

application dated April 29, 1996, as

supplemented October 21, December 2,

and December 16, 1996, and reviewed

in the staff’s safety evaluation dated

January 8, 1997.

Amendment Nos.: Unit 1 - 170, Unit

2 - 174

Facility Operating License Nos. DPR24 and DPR-27: Amendments revise the

Technical Specifications and the

operating licenses.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28621)

The October 21, December 2, and

December 16, 1996, supplements

provided corrected license and TS pages

and a 90-day implementation schedule.

These supplements were within the

scope of the original application and did

not change the staff’s initial proposed

no significant hazards considerations

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

January 8, 1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

October 31, 1996
4360 Federal Register / Vol. 62, No. 19 / Wednesday, January 29, 1997 / Notices

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant Technical Specification 6.9,

‘‘Reporting Requirements,’’ by deleting

the annual requirement to submit a

description of changes made pursuant to

10 CFR 50.59. Administrative changes

are also made to correct inconsistencies

in the TS Table of Contents and in a

footnote for Table TS 3.5-1.

Date of issuance: January 6, 1997

Effective date: January 6, 1997

Amendment No.: 131

Facility Operating License No. DPR43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64397) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 6, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Dated at Rockville, Maryland, this 22nd

day of January 1997.

For The Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 97–1994 Filed 1-28-97; 8:45 am]

BILLING CODE 7590-01-F

[Docket Nos. 50–255 and 72–7]

Consumers Power Co., Palisades

Nuclear Plant, License Nos. DPR–20;

Issuance of Director’s Decision Under

10 CFR 2.206

Notice is hereby given that the Acting

Director, Office of Nuclear Reactor

Regulation, has issued a Director’s

Decision concerning a Petition dated

September 19, 1995, as amended on

September 30, 1996, filed by Don’t

Waste Michigan and Lake Michigan

Federation (Petitioners) under Section

2.206 of Title 10 of the Code of Federal

Regulations (10 CFR 2.206). The Petition

requested that the NRC (1) find that

Consumers Power Company (licensee)

violated NRC requirements related to

unloading procedures for dry storage

casks for spent nuclear fuel, (2) suspend

the licensee’s use of the general license

provisions related to dry cask storage of

spent nuclear fuel, (3) require a

substantial penalty be paid by the

licensee, and (4) conduct hearings

related to unloading procedures for dry

storage casks at Palisades.

The Acting Director of the Office of

Nuclear Reactor Regulation has

determined that Petition should be

granted in part and denied in part for

the reasons stated in the ‘‘Director’s

Decision Under 10 CFR 2.206’’ (DD–97–

01), the complete text of which follows

this notice. The decision and documents

cited in the decision are available for

public inspection and copying in the

Commission’s Public Document Room,

the Gelman Building, 2120 L Street,

NW, Washington, DC, and at the local

public document room located in the

Van Wylen Library at Hope College in

Holland, Michigan.

A copy of this decision has been filed

with the Secretary of the Commission

for the Commission’s review in

accordance with 10 CFR 2.206(c). As

provided therein, this decision will

become the final action of the

Commission 25 days after issuance

unless the Commission, on its own

motion, institutes review of the decision

within that time.

Dated at Rockville, MD., this 23d day of

January 1997.

For the Nuclear Regulatory Commission.

Frank J. Miraglia,

Acting Director, Office of Nuclear Reactor

Regulation.

Director’s Decision Under 10 CFR 2.206

I. Introduction

On September 19, 1995, the

organizations Don’t Waste Michigan and

Lake Michigan Federation (Petitioners)

filed a Petition pursuant to Section

2.206 of Title 10 of the Code of Federal

Regulations (10 CFR 2.206) requesting

that the U.S. Nuclear Regulatory

Commission (NRC) (1) find that

Consumers Power Company (licensee)

violated NRC requirements related to

unloading procedures for dry storage

casks for spent nuclear fuel, (2) suspend

the licensee’s use of the general license

provisions related to dry cask storage of

spent nuclear fuel, (3) require a

substantial penalty be paid by the

licensee, and (4) conduct hearings

related to unloading procedures for dry

storage casks at Palisades.

On September 30, 1996, the

Petitioners amended the Petition by

including additional information in

support of their position that the

licensee did not have a workable

unloading procedure before loading the

13 dry storage casks currently in the

Palisades independent spent fuel

storage installation (ISFSI).

The Petition has been referred to me

pursuant to 10 CFR 2.206. The NRC

letter dated October 24, 1995, to Dr.

Sinclair and Mr. Skavroneck, on behalf

of the Petitioners, acknowledged receipt

of the Petition. Notice of receipt was

published in the Federal Register on

October 31, 1995 (60 FR 55388).

On the basis of the NRC staff’s

evaluation of the issues and for the

reasons given below, the Petitioners’

requests are granted in part and denied

in part.

II. Background

NRC regulations contain a general

license that authorizes nuclear power

plants licensed by the NRC, such as

Palisades, to store spent nuclear fuel at

the reactor site in storage casks

approved by the NRC. (See 10 CFR part

72, subpart K.) In regard to dry cask

storage of spent nuclear fuel at

Palisades, the licensee opted to use the

VSC–24 Cask Storage System designed

by Sierra Nuclear Corporation. The

VSC–24 Cask Storage System was added

to the list of NRC certified casks in May

1993 (58 FR 17948). The associated

certificate of compliance, Certificate

Number 1007, specifies the conditions

for use of VSC–24 casks under the

general license provisions of 10 CFR

part 72. Section 1.1.2, ‘‘Operating

Procedures,’’ in the certificate of

compliance for the VSC–24 casks,

requires that licensees prepare an

operating procedure related to cask

unloading. Specifically, the condition

states

Written operating procedures shall be

prepared for cask handling, loading,

movement, surveillance, and maintenance.

The operating procedures suggested

generically in the SAR (safety analysis report)

are considered appropriate, as discussed in

Section 11.0 of the SER (safety evaluation

report), and should provide the basis for the

user’s written operating procedures. The

following additional written procedures shall

also be developed as part of the user

operating procedures:

1. A procedure shall be developed for cask

unloading, assuming damaged fuel. If fuel

needs to be removed from the multi-assembly

sealed basket (MSB), either at the end of

service life or for inspection after an

accident, precautions must be taken against

the potential for the presence of oxidized fuel

and to prevent radiological exposure to

personnel during this operation. This activity

can be achieved by the use of the Swagelok

valves, which permit a determination of the

atmosphere within the MSB before the

removal of the structural and shield lids. If

the atmosphere within the MSB is helium,

then operations should proceed normally,

with fuel removal, either via the transfer cask

or in the pool. However, if air is present

within the MSB, then appropriate filters

should be in place to permit the flushing of

any potential airborne radioactive particulate

from the MSB, via the Swagelok valves. This

action will protect both personnel and the

operations area from potential

contamination. For the accident case,

personnel protection in the form of

respirators or supplied air should be