Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66133

Information Collection Request

[paperwork package]. Comments

submitted in response to this notice will

be summarized and/or included in the

request for Office of Management and

Budget approval of the information

collection request; they will also

become a matter of public record.

Dated: December 9, 1997.

Adam M. Finkel,

Director, Directorate of Health Standards

Programs.

[FR Doc. 97–32863 Filed 12–16–97; 8:45 am]

BILLING CODE 4510–26–M

NATIONAL ARCHIVES AND RECORDS

ADMINISTRATION

Public Evaluation of NARA Archival

Information Locator (NAIL)

AGENCY: National Archives and Records

Administration (NARA).

ACTION: Notice.

SUMMARY: NARA is inviting the public

to participate in an evaluation of its

prototype online information system,

the NARA Archival Information Locator

(NAIL).

As part of its Electronic Access

Project, NARA is constructing a

nationwide, integrated online

information delivery system. The

project, a priority under the agency’s

Strategic Plan, will eventually result in

a virtual card catalog of all NARA

holdings nationwide, including those in

the Presidential libraries and regional

archives. In addition, copies of some of

NARA’s most popular and significant

manuscripts, photographs, sound

recordings, maps, drawings and other

documents will be digitized and

available for researchers to view online

through the catalog.

To complete the final functional

requirements for the catalog, NARA is

undertaking an evaluation of its

prototype, the NARA Archival

Information Locator (NAIL). All

members of the public are invited to use

NAIL and to comment on its ease of use,

functionality, and terminology.

NAIL can be accessed on the World

Wide Web at http://www.nara.gov/nara/

nail.html.

DATES: Comments should be received by

January 31, 1998.

ADDRESSES: Comments can be sent

through the online comments link in

NAIL or by e-mail to

nail.mailbox@arch2.nara.gov.

Dated: December 10, 1997.

L. Reynolds Cahoon,

Assistant Archivist for Human Resources and

Information Services.

[FR Doc. 97–32914 Filed 12–16–97; 8:45 am]

BILLING CODE 7515–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from November

21, 1997, through December 5, 1997.

The last biweekly notice was published

on December 3, 1997 (62 FR 63970).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By January 16, 1998, 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
66134 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

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.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois, Docket Nos. STN

50–456 and STN 50–457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of amendment request: February

28, 1997.

Description of amendment request:

The proposed amendments would

revise Byron and Braidwood Technical

Specifications (TS) Sections 3/4.4.5,

‘‘Steam Generators,’’ and 3/4.4.8,

‘‘Reactor Coolant System Specific

Activity,’’ for both the Byron Station,

Units 1 and 2, and the Braidwood

Station, Units 1 and 2. The intent of

these proposed revisions is to restore for

both Byron, Unit 1, and Braidwood,

Unit 1, the original TS related to steam

generator (SG) inspections and the

primary coolant dose equivalent iodine131 (DEI) concentrations. These

amendments will become effective

when the original steam generators

(OSG) which are Westinghouse Model

D4 SGs, are removed and the

replacement steam generators (RSG)

made by Babcock and Wilcox,

International (BWI), are installed. The

RSGs are presently being installed at

Byron, Unit 1, while the RSGs will be

installed at Braidwood, Unit 1, in fall

1998.

The SG inspection methodology,

inspection frequency, reporting

requirements and acceptance criteria for

the RSGs in both Byron, Unit 1, and

Braidwood, Unit 1, will revert to the

TSs for the OSGs before several prior

license amendments incorporated into

the TSs: (1) The interim plugging

criteria (IPC) consistent with Generic

Letter (GL) 95–05; (2) the F* criteria for

the SG tube expansions into the

tubesheet; and (3) the criteria for

repairing SG tubes using either

Westinghouse laser welded sleeves or

Combustion Engineering tungsten inert

gas (TIG) welded sleeves. The TSs

applicable to Byron, Unit 2, and

Braidwood, Unit 2, both of which have

Westinghouse Model D5 SGs, remain

unchanged except for designating them

in the TSs as model D5 SGs.

With respect to the limiting value of

the DEI primary coolant concentration,

both the Byron, Unit 1, TSs and the

Braidwood, Unit 1, TSs will revert from

their present TS limit of 0.35 to 1.0

microcuries per gram. A license

amendment request to lower the Byron,

Unit 1, TS DEI limit from 0.35 to 0.20

microcuries per gram was submitted on

January 31, 1997, but this request was
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66135

subsequently withdrawn on November

11, 1997, because the RSGs were being

installed in the Byron, Unit 1, refueling

outage which started in early November

1997. A license amendment request to

lower the Braidwood, Unit 1, TS DEI

limit from 0.35 to 0.10 microcuries per

gram was submitted on September 2,

1997. Action on this request is still

pending but in any case, will not affect

the subject license amendment request

for Braidwood, Unit 1, because the

September 2, 1997, request is only

applicable to the OSGs which are

presently using the IPC that were

originally incorporated into the TSs on

November 9, 1995. The applicable bases

sections of the Byron, Unit 1, TSs and

Braidwood, Unit 1, TS will also be

revised to reflect the TS changes

discussed above.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Due to design differences between the

replacement Steam Generators (RSGs) and

OSGs, the analyses supporting the

application of the F* and voltage-based

repair criteria do not apply to the RSGs. Also,

the analyses supporting sleeving repair by

the Westinghouse laser welded or

Combustion Engineering Tungsten Inert Gas

(TIG) welded sleeving methodologies do not

apply to the RSGs due to the design

differences. The RSG and OSG tube bundle

configurations are similar, however, the RSG

tubes are smaller in diameter, constructed of

Inconel Alloy 690 instead of Alloy 600, and

supported by stainless steel lattice grids

instead of the drilled carbon steel plates used

in the OSGs. The RSG tubes are hydraulically

expanded into the tube sheet during initial

assembly. The RSG upper tube bundle shape

consists of tubes with continuous, smooth,

long radius bends.

The structural analysis demonstrates that

the tube integrity is maintained for a Main

Steamline Break (MSLB) occurring during

normal full power operation. The structural

evaluation of the tubing for faulted

conditions was performed in accordance

with the ASME Boiler and Pressure Vessel

Code Section III requirements. The tube

material selection and size exceed the

strength requirements of the existing steam

generators. Comparison of the Alloy 690 tube

material used in the RSGs with the Alloy 600

tube material in the OSGs show that the RSG

material strength characteristics are as good

as or better than those of the existing design.

A comparison of the stress margins of the

RSG and OSG show that the stress margin in

the RSG tubes exceed the stress margin in the

OSG tubes.

RSG portions of the reactor coolant

pressure boundary are designed to permit

periodic inspection and testing of important

areas and features to assess structural and

leak-tight integrity. ASME Section XI,

provides the depth of an allowable outside

diameter (O.D.) flaw for tubes in service. The

RSG has tubing fabricated from SB–163

material (Inconel Alloy 690) which is

examined by eddy current methods to the

requirements of ASME Section III, NB–2550.

The tubing has a radius to thickness (r/t) ratio

less than 8.70. In accordance with ASME

Section XI, for tubing having an r/t ratio of

less than 8.70, the depth of an allowable O.D.

flaw shall not exceed 40% of the nominal

tube wall thickness.

The potential for tube rupture is not

increased from the OSGs as demonstrated in

the qualification analysis and testing for the

RSGs. The program for periodic inservice

inspection of the steam generators monitors

the integrity of the SG tubing to ensure that

there is sufficient time to take proper and

timely corrective action if any tube

degradation is detected. Therefore,

installation of the RSGs will not increase the

probability of the occurrence of primary-tosecondary leakage or a steam generator tube

rupture (SGTR) during normal or accident

conditions.

The design basis doses calculated for

postulated accidents involving degradation of

SG tubes, such as SGTR and MSLB accidents,

as presented in UFSAR [Updated Final Safety

Analysis Report] Chapter 15 accident

analysis have been evaluated and are

decreased by installation of the RSGs and

restoration of the RCS activity limit to 1.0

microcuries/gm. The decrease in offsite dose

is primarily due to the smaller RSG tube

diameter and less primary-to-secondary

transfer during the event. The dose

calculations are performed consistent with

NUREG–0800, ‘‘Standard Review Plan’’ and

ensure site boundary doses are within a small

fraction of the Title 10 Code of Federal

Regulations Part 100 (10 CFR 100)

requirements. Therefore, the change does not

involve a significant increase in the

consequences of an accident previously

evaluated.

Limiting the applicability of TS provisions

to a specific cycle or SG type are

administrative changes in that they provide

clarification consistent with current analyses

and do not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Restricting application of IPC, F* and

sleeving methodologies to the OSGs and

reinstating an RCS activity limit of 1.0

microcuries/gm upon installation of the RSGs

will not introduce significant or adverse

changes to the plant design basis that could

lead to a new or different kind of accident

being created. The RSG tubing meets the

requirements of General Design Criteria

(GDC) 14, 15, 30, 31, and 32 of 10 CFR 50,

Appendix A. The RSG tubing has been

designed and evaluated consistent with

ASME Code Section III criteria and the

inspection criteria for the RSGs is consistent

with ASME Code Section XI criteria. The

RSGs have thermally treated Inconel Alloy

690 tubes which are hydraulically expanded

into the tube sheet during initial assembly.

Alloy 690 is more resistant to stress corrosion

cracking (SCC) than Alloy 600 which is used

in the OSG tubing. Overall tube bundle

structural and leakage integrity is maintained

at a level consistent with or better than the

originally supplied tubing during all plant

conditions.

ComEd will continue to apply the TS

maximum primary-to-secondary leakage limit

of 150 gpd (0.1 gpm) through any one SG at

Byron and Braidwood to help preclude the

potential for excessive leakage during all

plant conditions. The EPRI recommended

150 gpd limit provides for leakage detection

and plant shutdown in the event of an

unexpected tube leak and precludes the

potential for excessive leakage or tube burst

in the event of a Main Steam Line Break or

under Loss of Coolant Accident conditions.

Limiting the applicability of TS provisions

to a specific cycle or SG type are

administrative changes in that they provide

clarification consistent with current analyses.

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 change does not involve

a significant reduction in a margin of safety.

Restricting application of IPC, F*, and

sleeving methodologies to the OSGs for

which the supporting analyses apply, does

not involve a reduction in a margin of safety.

The RSG tubing has been shown to retain

adequate structural and leakage integrity

during normal, transient, and postulated

accident conditions consistent with GDC 14,

15, 30, 31, and 32 of 10 CFR 50 Appendix

A. The RSG tubing has been designed and

evaluated consistent with the margins of

safety specified in ASME Code Section III.

The proposed program for periodic inservice

inspection of the replacement steam

generators monitors the integrity of the SG

tubing to ensure that there is sufficient time

to take proper and timely corrective action if

any tube degradation is present. The

proposed program is consistent with the

Standard Technical Specifications.

The Unit 1 RCS dose equivalent I–131 limit

is being raised upon installation of the RSGs

to eliminate the compensatory lower limit

that was adopted in conjunction with IPC for

the existing Westinghouse D4 SGs. With the

RCS activity limit returned to the Standard

Technical Specification value of 1.0 [mu]Ci/

gm, the assessment of postulated UFSAR

Chapter 15 accidents (including SGTR and

MSLB) has concluded that the calculated

design basis doses presented in Chapter 15

are not adversely impacted by the RSGs. This

ensures that the resulting 2-hour dose rates

at the Byron and Braidwood site boundaries

will not exceed an appropriately small

fraction of 10 CFR 100 dose guideline values.

Limiting the applicability of TS provisions

to a specific cycle or SG type are
66136 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

administrative changes in that they provide

clarification consistent with current analyses.

Therefore, it is concluded that this change

does not involve a significant reduction in a

margin of safety with respect to plant safety

as defined in the UFSAR or the Technical

Specification.

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: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Robert A. Capra.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: March

26, 1997.

Description of amendment request:

The proposed amendment would revise

the containment system technical

specifications (TS) contained in TS

Sections 3.6 and 4.5. The licensee has

classified the changes as ‘‘More

Restrictive,’’ ‘‘Less Restrictive,’’ and

‘‘Administrative.’’ ‘‘More Restrictive’’

changes include reduction of the

allowable containment pressure,

addition of an action statement defining

action to be taken when the

containment pressure limit is exceeded,

addition of a restriction on containment

temperature, and revision of the

applicable conditions for the

containment purge valves to require that

the valves be operable above 210

degrees F versus the current

requirement that they be operable above

525 degrees F. ‘‘Less Restrictive’’

changes include addition of an

allowance to enter an air lock through

a locked door to perform maintenance,

addition of an allowance to open

containment isolation valves under

administrative control, revision of the

applicable conditions for containment

pressure to exclude the cold shutdown

operating condition, and addition of an

exception to the surveillance

requirement requiring verification of the

status of ‘‘locked-closed’’ manual

isolation valves after a refueling outage

to exclude requiring such verification

for valves opened under administrative

control. ‘‘Administrative’’ changes

include the deletion of containment

isolation valve tables and component

identifiers from the TS in accordance

with Generic Letter 91–08 (‘‘Removal of

Component Lists from Technical

Specifications’’) and editorial

restructuring of the affected TS sections

to clarify the remaining requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Each proposed change has been classified

as ‘‘Administrative,’’ ‘‘More Restrictive,’’ or

‘‘Less Restrictive.’’ ‘‘Administrative’’ and

‘‘More Restrictive’’ changes are discussed

generically; ‘‘Less Restrictive’’ changes are

discussed individually.

Five of the proposed changes are classified

as being ‘‘Less Restrictive’’:

(G.1) Allowance in LCO [Limiting

Condition for Operation] 3.6.1 to enter an air

lock to perform maintenance.

(G.2) Allowance in LCO 3.6.1 to open

containment isolation valves under

administrative control.

(I.2) Revising the applicable conditions of

LCO 3.6.2, Containment Pressure to exclude

Cold Shutdown.

(J.2) Exception in SR [Surveillance

Requirement] 4.5.3d for valves opened under

administrative control as allowed by LCO

3.6.1.

(P) Allowance in SR 4.5.2 to enter an air

lock to perform maintenance.

Four of the proposed changes are classified

as being ‘‘More Restrictive’’:

(I.1) Revising LCO 3.6.2 to reduce the

allowable containment pressure.

(I.3) Addition of an action statement to

LCO 3.6.2, Containment Pressure.

(K) Addition of a new LCO which restricts

Containment Temperature.

(M.2) Revising the applicable conditions

for LCO 3.6.5, Purge Valves.

The remaining changes are all classified as

being ‘‘Administrative’’.

Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

1. Changes G.1, G.2, J.2, and P: Proposed

changes G.1 and P allow limited access

through the operable door of an air lock

when the other door is inoperable; current

Technical Specifications [TS] do not.

Proposed changes G.2 and J.2 allow

unisolating containment penetration flow

paths intermittently under administrative

control; current TS do provide a similar

allowance, but only for one specific

penetration. These changes cannot

significantly increase the probability of an

accident because opening an air lock door or

a containment penetration is not, itself, an

initiator and does not affect the items which

are initiators of any analyzed accident.

The ability to open the operable door or to

open a containment penetration, even if it

means the containment boundary is

temporarily not intact, does not significantly

increase the consequences of an accident

previously evaluated because of the low

probability of an event that could pressurize

the containment occurring during the short

time the operable door or containment

penetration is expected to be open. In a case

where containment integrity (or containment

operability) is lost due to excessive leakage,

both the Palisades Technical Specifications

and the Standard Technical Specifications

[STS] allow one hour of continued operation

for its restoration. That time period is

allowed without regard to the magnitude of

the potential leakage, and would be allowed

even if both personnel air lock doors [were]

leaking excessively. The additional

allowance of permitting the operable door to

be opened momentarily for entry or egress

when the other door is inoperable due to

excessive leakage would not significantly add

to the probability of containment leakage and

the resultant consequences of an accident.

Similarly, the allowance to open any

containment penetration intermittently under

administrative control, which currently is

allowed for one penetration, would not

significantly add to the probability of

containment leakage and the resultant

consequences of an accident.

Therefore, operation of the Facility in

accordance with proposed changes G.1, G.2,

J.2, and P would not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Change I.2: Change I.2 alters existing

LCO 3.6.2, Containment Pressure so that it no

longer applies during Cold Shutdown. LCO

3.6.2 is intended to limit containment

pressure to that value used as an initial

condition in the safety analysis. Containment

pressure is an initial condition in analyses

which assure that containment internal

pressure will not exceed the containment

design values during a LOCA or MSLB.

Containment pressure is not an initiator of

any accident previously evaluated. Neither a

LOCA [loss-of-coolant accident] nor a MSLB

[main steam line break] occurring during

Cold Shutdown would pressurize the

containment. Therefore, a containment

pressure LCO is not necessary, during Cold

Shutdown, to assure that containment design

pressure and temperature is not exceeded.

The STS Containment pressure LCO is not

applicable in Cold Shutdown.

Therefore, operation of the Facility in

accordance with proposed change I.2 would

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

3. More Restrictive Changes: ‘‘More

Restrictive’’ changes only add new

requirements, or revise existing requirements

to result in additional operational

restrictions. The TS, with all ‘‘More

Restrictive’’ changes incorporated, will still

contain all of the requirements which existed

prior to the changes. Therefore, ‘‘More

Restrictive’’ changes cannot involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

4. ‘‘Administrative’’ changes make wording

changes which clarify existing TS

requirements, without affecting their
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66137

technical content. Since ‘‘Administrative’’

changes do not alter the technical content of

any requirements, they cannot involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Do the proposed changes create the

possibility of a new or different kind of

accident from any previously evaluated?

1. Changes G.1, G.2, J.2, and P: Proposed

changes G.1 and P allow limited access

through the operable door of an air lock

when the other door is inoperable; current

Technical Specifications do not. Proposed

changes G.2 and J.2 allow unisolating

containment penetration flow paths

intermittently under administrative control;

current TS do provide a similar allowance,

but only for one specific penetration.

Opening an air lock door or a containment

penetration does not affect the operating

conditions or operation of any plant systems

(other than the containment); it does not

create a threat to the integrity of any

operating system or alter any system

operating practice or settings.

Since the opening of an air lock door or a

containment penetration only affects the

potential leakage from the containment, and

does not affect any of the operating plant

systems, operation of the Facility in

accordance with the proposed Technical

Specifications change would not create the

possibility of a new or different kind of

accident from any previously evaluated.

2. Change I.2: Change I.2 alters existing

LCO 3.6.2, Containment Pressure so that it no

longer applies during Cold Shutdown. LCO

3.6.2 is intended to limit containment

pressure to that value used as an initial

condition in the safety analysis. Containment

pressure is an initial condition in analyses

which assure that containment internal

pressure will not exceed the containment

design values during a LOCA or MSLB.

Neither a LOCA nor a MSLB occurring

during Cold Shutdown would pressurize the

containment. Therefore, a containment

pressure LCO is not necessary, during Cold

Shutdown, to avoid creation of a new or

different kind of accident. The STS

Containment pressure LCO is not applicable

in Cold Shutdown.

Therefore, operation of the Facility in

accordance with proposed change I.2 would

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. More Restrictive Changes: ‘‘More

Restrictive’’ changes only add new

requirements, or revise existing requirements

to result in additional operational

restrictions. The TS, with all ‘‘More

Restrictive’’ changes incorporated, will still

contain all of the requirements which existed

prior to the changes. Therefore, ‘‘More

Restrictive’’ changes cannot create the

possibility of a new or different kind of

accident from any previously evaluated.

4. ‘‘Administrative’’ changes make wording

changes which clarify existing TS

requirements, without affecting their

technical content. Since ‘‘Administrative’’

changes do not alter the technical content of

any requirements, they cannot create the

possibility of a new or different kind of

accident from any previously evaluated.

Do the proposed changes involve a

significant reduction in a margin of safety?

1. Changes G.1, G.2, J.2, and P: Proposed

changes G.1 and P allow limited access

through the operable door of an air lock

when the other door is inoperable; current

Technical Specifications do not. Proposed

changes G.2 and J.2 allow unisolating

containment penetration flow paths

intermittently under administrative control;

current TS do provide a similar allowance,

but only for one specific penetration. The

ability to open the operable door or a

containment penetration, even if it means the

containment boundary is temporarily not

intact, does not involve a significant

reduction in a margin of safety because of the

low probability of an event that could

pressurize the containment occurring during

the short time the operable door or

penetration is expected to be open.

Therefore, operation of the Facility in

accordance with the proposed Technical

Specifications change would not involve a

significant reduction in a margin of safety.

2. Change I.2: Change I.2 alters existing

LCO 3.6.2, Containment Pressure so that it no

longer applies during Cold Shutdown. LCO

3.6.2 is intended to limit containment

pressure to that value used as an initial

condition in the safety analysis. Containment

pressure is an initial condition in analyses

which assure that containment internal

pressure will not exceed the containment

design values during a LOCA or MSLB.

Neither a LOCA nor a MSLB occurring

during Cold Shutdown would pressurize the

containment. Therefore, elimination of a

Cold Shutdown LCO for containment

pressure would not affect the post-accident

pressure or temperature. Since peak post

accident [pressure] and temperature would

be unaffected by the proposed change,

operation of the Facility in accordance with

proposed change I.2 would not involve a

significant reduction in a margin of safety.

3. More Restrictive Changes: ‘‘More

Restrictive’’ changes only add new

requirements, or revise existing requirements

to result in additional operational

restrictions. The TS, with all ‘‘More

Restrictive’’ changes incorporated, will still

contain all of the requirements which existed

prior to the changes. Therefore, ‘‘More

Restrictive’’ changes cannot involve a

significant reduction in a margin of safety.

4. ‘‘Administrative’’ changes make wording

changes which clarify existing TS

requirements, without affecting their

technical content. Since ‘‘Administrative’’

changes do not alter the technical content of

any requirements, they cannot 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Energy Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: John N.

Hannon.

Carolina Power & Light Company, et

al., Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of amendments request: October

29, 1997.

Description of amendments request:

The proposed amendments to the

Technical Specifications (TS) for the

Brunswick Steam Electric Plant (BSEP)

Units 1 and 2 would revise the

description of the control rod

assemblies (CRAs) in TS 5.3.2. The

proposed revision was requested to

support replacement of a portion of the

BSEP Unit 1 CRAs during that unit’s

next refueling outage with assemblies of

a different design. Carolina Power &

Light Company, the licensee, has

proposed adopting the description of

CRAs used in NUREG–1433, Revision 1,

‘‘Standard Technical Specifications

General Electric Plants, BWR/4,’’ which

includes the number and shape of CRAs

and a stipulation that NRC-approved

absorber material be used in CRAs. The

more detailed description in the current

TS of CRAs would be relocated to the

Updated Final Safety Analysis Report.

The licensee has stated that the CRA

description proposed for TS 5.3.2 will

be sufficient to ensure that any future

changes in CRA design that may affect

safety will require prior NRC review and

approval.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendments do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Relocation of the control rod assembly

descriptive information from the Technical

Specifications to the Updated Final Safety

Analysis Report will ensure that adequate

control of the information is maintained. Any

changes to this design information must

conform with the requirements of 10 CFR

50.59. Restricting use of control rod assembly

absorber materials to those listed, or to

materials that have been approved by the

NRC, will ensure any changes which may

affect safety to require prior NRC review and

approval. Since the information with a

potential to affect safety is sufficiently

addressed by the Technical Specifications,

the criteria of 10 CFR 50.36(c)(4) for
66138 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

including the relocated information as Design

Features are not met. Because the relocated

information is not required to be in the

Technical Specifications to provide adequate

protection of the public health and safety,

relocation of control rod assembly

descriptive information will not increase

either the probability or the consequences of

an accident previously evaluated.

2. The proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Relocation, to the Updated Final Safety

Analysis Report, of the information

pertaining to the control rod assembly

designs ensures that adequate control of the

information will be maintained. Since the

information with a potential to affect safety

is sufficiently addressed by the Technical

Specifications, the criteria of 10 CFR

50.36(c)(4) for including the relocated

information as Design Features are not met.

Because the relocated information is not

required to be in the Technical Specifications

to provide adequate protection of the public

health and safety, the proposed Technical

Specification changes to relocate the control

rod assembly design information to the

Updated Final Safety Analysis Report does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety.

As discussed in Items 1 and 2 above,

relocation of the control rod assembly

descriptive information from the Technical

Specifications to the Updated Final Safety

Analysis Report will ensure that adequate

control of the information is maintained. Any

changes to this design information must

conform with the requirements of 10 CFR

50.59. Restricting use of control rod assembly

absorber materials to those listed, or to

materials that have been approved by the

NRC, will ensure any changes which may

affect safety to require prior NRC review and

approval. The information with a potential to

affect safety is sufficiently addressed by the

Technical Specifications, therefore, the

proposed Technical Specification changes to

relocate control rod assembly design

information to the Updated Final Safety

Analysis Report 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: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297

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: James E. Lyons.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: July 16,

1997, as supplemented October 30,

1997.

Description of amendment request:

The amendment would update License

condition 2.C(4) to reflect the latest

revision levels of the Oyster Creek

Security Training and Qualification

Plan, License Amendment Request No.

252.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

GPU Nuclear has concluded that the

proposed changes to the Security Plan do not

involve a significant hazard consideration. In

support of this determination, an evaluation

of each of the three standards set forth in 10

CFR 50.92 is provided below.

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Security Plan provisions are not associated

with design basis accident initiators nor do

they constitute part of any mitigation system.

Therefore, the probability and consequences

of accidents 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 Security Plan changes do not create

new or change existing physical interfaces

with plant equipment. Therefore, the changes

do not create the possibility of a new or

different kind of accident.

(3) The proposed changes do not involve

a significant reduction in a margin of safety.

Margins associated with reactor and fuel

storage nuclear safety are not affected by the

proposed Security Plan changes since neither

physical nor procedural changes to

associated systems, structures and

components are involved. Vital area security

measures, which are reduced, are

compensated by commitments to hold

contingency drills at a frequency sufficient to

maintain response capability for response

personnel and to use organic-type X-ray

equipment.

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: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Ronald B.

Eaton, Acting Director.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request:

November 14, 1997.

Description of amendment request:

The proposed change to Technical

Specification 4.5.2.d.1 will clarify the

wording and increase the setpoint for

the open pressure interlock (OPI).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

revision in accordance with 10 CFR 50.92

and has concluded that the revision does not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does not

involve [an] SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

Increasing the Technical Specification

Open Pressure Interlock (OPI) pressure to

412.5 psia [pounds per square inch—

atmospheric] will still maintain the required

function of preventing the MOVs [motor

operated valves] from opening inadvertently.

The increased pressure is within the design

limits of the RHR [residual heat removal]

piping system and components. The pressure

signal is generated from a transmitter and

results in an electronic input to the bistable.

This is a clarification of the conditions under

which the OPI is tested.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

There is no change to the function of the

OPI. The protection provided by the interlock

remains intact. The Technical Specification

OPI pressure has been raised to take into

account instrument accuracies and reset

deadbands. The RHR system design pressure

remains protected from being exceeded by

inadvertent opening of the isolation MOVs.

The method for the OPI surveillance is

clarified by clearly stating that the bistable

receives a simulated transmitter signal

representative of the process pressure.
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66139

Therefore, the proposed revision does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The design pressure of the RHR system is

600 psig [pounds per square inch—gauge].

The most limiting case is to prevent the RHR

pump developed head pressure from

exceeding the design pressure when aligned

to the RCS [reactor coolant system] as suction

pressure. RHR pump testing has determined

that a maximum pump differential pressure

of 195 psi [pounds per square inch] exists for

deadhead/no flow conditions. Therefore, to

maintain the 600 psig design pressure limit,

RCS/suction pressure must be limited to 405

psig (420 psia, assuming a 15 psi conversion

from psig to psia). The proposed maximum

pressure, including setpoint tolerances and

reset deadbands, is less than this value; i.e.

412.5 psia. Head corrections due to elevation

differences are considered to be insignificant.

Therefore, the proposed revision does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

revision does not involve an SHC.

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: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request:

November 3, 1997.

Description of amendment request:

The proposed amendment would revise

Sections 1, 3.1, 3.3, 4.3, and 6 of

Appendix A of the Indian Point 3

Technical Specifications. These

revisions extend the Heatup-Cooldown

limits from 11 to 13 effective full power

years (EFPYs), provide the

corresponding Overpressure Protection

System (OPS) limits, relocate the new

pressure temperature limit curves and

low-temperature overpressurization

protection (LTOP) system limits to the

pressure temperature limit report

(PTLR) and include some minor

revisions which ensure specification

clarity and conservatism.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously analyzed?

Response: The proposed license

amendment does not involve a significant

increase in the probability or consequences

of a previously analyzed accident. The

pressure-temperature limit changes proposed

by this amendment are based on supporting

data and evaluation methodologies

previously submitted to the NRC in

Reference 3 [see application dated November

3, 1997] and approved as Amendments 109

and 121 (References 4 and 5) [see application

dated November 3, 1997]. These limits are

based upon the irradiation damage prediction

methods of Regulatory Guide 1.99, Revision

2. The LTOPs changes contained in this

submittal have been conservatively adjusted

in accordance with the new pressuretemperature limits, in accordance with the

methodology contained in Reference 3 and

ASME Code Case N–514.

The relocation of the pressure-temperature

and LTOPs limits from the Technical

Specifications to the PTLR does not eliminate

the requirement to operate in accordance

with the limits specified in 10 CFR [Part] 50,

Appendix G. The requirement to operate

within the limits in the PTLR is specified in

and controlled by the Technical

Specifications.

The revised version of Section 3.1.A.8

clarifies existing requirements related to the

OPS system and adds an eight hour

completion time for compensating actions,

consistent with the STS [standard technical

specifications]. The changes to Section

3.1.A.1.h, i, and j revise the requirements

associated with the start of an RCP [reactor

coolant pump]. These changes improve

specification clarity and do not increase the

probability or consequences of an accident.

The Technical Specification changes

associated with the restriction on SI [safety

injection] pumps provides added

conservatism to the Technical Specifications

and limits the likelihood of an RHR [residual

heat removal] overpressurization event.

Current plant procedures prohibit actuation

of any SI pumps when RHR is in service,

except during testing, loss of RHR cooling, or

reduced inventory operations. Therefore, the

change to the Technical Specifications will

not alter current plant operation.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response: The proposed license

amendment does not create the possibility of

a new or different kind of accident from any

accident previously analyzed. The pressuretemperature limits are updating the existing

limits by taking into account the effects of

radiation embrittlement, utilizing criteria

defined in Regulatory Guide 1.99, Revision 2,

and extending the effective period to 13

EFPYs. The updated OPS limits have been

adjusted to account for the effect of

irradiation on the limiting reactor vessel

material. These changes do not affect the way

the pressure-temperature or OPS limits

provide plant protection and no physical

plant alterations are necessary. The

relocation of the pressure-temperature and

OPS limits from the Technical Specifications

to the PTLR does not alter the requirements

associated with these limits.

The revisions to Section 3.1.A.8

concerning the OPS system improve on the

clarity of existing specifications and add a

completion time for compensating actions

that is consistent with the STS. These

changes do not involve any hardware

modifications and do not affect the function

of the OPS system.

The revisions concerning the operation of

SI pumps bring the Technical Specifications

into line with current operating procedures.

The changes to Specification 3.1.A.1.h, i, and

j provide specification clarity and are more

conservative than existing Technical

Specifications. Therefore, the changes cannot

create the possibility of a new or different

kind of accident.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response: The proposed amendment does

not involve a significant reduction in a

margin of safety. The margins of safety

against fracture provided by the pressuretemperature limits are those limits specified

in 10 CFR Part 50, Appendix G and ASME

Boiler and Pressure Vessel Code Section XI,

Appendix G. The guidance in these

documents has been utilized to develop the

pressure-temperature limits with the

requisite margins of safety for the heatup and

cooldown conditions. The new LTOP limits

are based upon Reference 3 and ASME Code

Case N–514. The relocation of the pressuretemperature and OPS limits to the PTLR does

not alter the requirements associated with

these limits.

The revisions to Section 3.1.A.8 clarify the

requirements associated with the OPS

system. The revisions associated with the

operation of SI pumps with RHR in service

(Sections 3.3.A.8, 9 and 10) and the changes

regarding RCP starts (Section 3.1.A.1.h, i, and

j) are more conservative than the current

Technical Specifications, and are consistent

with plant operating procedures. Therefore,

they do not reduce a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,
66140 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. David

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa.

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: October

24, 1997.

Description of amendment request:

The amendments would increase the

containment hydrogen analyzer

surveillance frequency in Technical

Specification 4.6.4.1 from once per

refueling outage to quarterly.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The containment hydrogen analyzers

provide control room indication of hydrogen

concentration in the containment

atmosphere. They do not affect the

probability of any previously evaluated

accident. The proposed change would

increase the calibration frequency specified

in TS 4.6.4.1 to make it consistent with

manufacturer’s recommendations and the

current calibration frequency at [Salem

Generating Station] SGS as imposed by

administrative controls. The change in TSrequired calibration frequency is in the

conservative (more frequent) direction, to

ensure that potential degradation of the

sensor electrolyte over time would not result

in unacceptable performance of the hydrogen

analyzers. The change in specified frequency

would not adversely affect the consequences

of any previously evaluated accident.

2. Proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

analyzed [evaluated].

The proposed change affects only the

specified calibration frequency of the

containment hydrogen analyzers. The

proposed change does not affect the design

of any SGS structure, system or component,

nor would it result in any new plant

configuration. Therefore, it does not create

the possibility of a new or different kind of

accident.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change to the containment

hydrogen analyzer calibration frequency does

not affect the design or operating limits of

any SGS structure, system or component. The

change would make the specified calibration

frequency more conservative, to ensure the

hydrogen analyzers perform as designed over

time. The proposed change does not involve

a significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

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: October

24, 1997.

Description of amendment request:

The amendments would revise

Technical Specification (TS) 3/4.7.7,

‘‘Auxiliary Building Exhaust Air

Filtration System.’’ The revisions

would: (1) Require both Auxiliary

Building Ventilation (ABVS) supply

fans to be operable, (2) require all three

ABVS exhaust fans to be operable, (3)

align ABVS TSs to be consistent with

current TS bases and recently revised

system descriptions in the Salem

Updated Final Safety Analysis Report

(UFSAR), (4) assure that negative

pressure is maintained in the Auxiliary

Building under all postulated single

active failures, (5) clarify required

Engineered Safety Feature filter testing,

(6) provide consistency between Unit 1

and Unit 2 TSs, and (7) for Unit 2 only,

remove the requirement to verify safety

injection auto-start capabilities.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change alters the number of

fans which must be OPERABLE to ensure

that a sufficient number of supply and

exhaust fans will be operable, following a

most limiting single failure, to mitigate the

consequences of design basis accidents. The

changes to the ABVS surveillance

requirements still provide an appropriate

means for demonstrating the operability of

the ABVS.

The ABVS cannot initiate or otherwise

cause any accident or operational transient

evaluated in the UFSAR. Consequently, the

probability of such events is not increased.

The ABVS cannot increase the consequences

of a design basis LOCA unless: (1) Auxiliary

Building negative pressure is lost, resulting

in uncontrolled, ground level release of

radioactive material; (2) ABVS carbon

adsorbers are bypassed, resulting in

uncontrolled release of radioactive iodine

from the plant vent; or (3) Auxiliary Building

temperatures are not controlled, resulting in

failure of accident mitigating equipment.

By requiring OPERABILITY of all ABVS

supply and exhaust fans, the proposed

changes contained in this submittal assures

Auxiliary Building negative pressure is

maintained under all postulated postaccident, single-failure scenarios. The

proposed changes to ABVS will not affect the

elemental iodine adsorption capability of the

system. Finally, engineering analyses

conclude that these fan combinations, with

single-active failures of the fans or their

support systems considered, provide

sufficient Auxiliary Building ventilation.

Under the most limiting temperature

conditions, the fans will maintain room

temperatures within design limits.

Accordingly, the consequences of a design

basis LOCA, hence applicable design basis

accidents or operational transients, are not

increased.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

ABVS supply fans are not considered

essential to the primary safety-function of

preventing or mitigating radioactive releases,

nor are they currently required to be

OPERABLE. Similarly, accident analyses take

no credit for operation of supply fans.

Accordingly, malfunctions of vital buses and

ABVS exhaust fans are the only malfunctions

of active ABVS related equipment important

to safety that are previously evaluated.

The probability of failure of a vital bus is

not increased by this proposal since the

proposal has no direct effect on electrical

power. Neither is the probability of exhaust

fan failure increased by the proposal, since

exhaust fans are not affected by this proposal,

except that the number that must be

OPERABLE is increased from two to three.

By requiring additional supply fans and

exhaust fans to be OPERABLE, no single

failure of either a vital bus or ABVS fan

prevents (1) maintenance of negative

Auxiliary Building pressure or (2)

maintenance of temperatures within design

limits. Since ABVS supply and exhaust fans

cannot initiate accidents, increasing the

number of fans required to be OPERABLE

cannot create the possibility of a new or

different kind of accident from any accident

previously evaluated. In addition, the

proposed changes to the ABVS surveillance

testing concern ABVS leakage, HEPA filter

and carbon adsorber capabilities, and

laboratory test methods. Therefore, the

proposed surveillance requirement changes

would have no impact on the initiation of

accidents.

Thus, the proposed changes do not create

the possibility of a new or different kind of
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66141

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety is dependent upon the

maintenance of specific operating parameters

within designated design limits. Since iodine

removal capability is not affected by the

proposed changes, and negative Auxiliary

Building pressure and temperatures will

continue to be maintained within existing

design limits under post-accident conditions,

including consideration of the most limiting

single active failure, the margin of safety is

not reduced. By imposing new restrictions on

the allowed outage times of ABVS

components, the margin of safety is increased

with the proposed changes to the ABVS

Technical Specification Limiting Condition

for Operation (LCO).

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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

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:

November 4, 1997.

Description of amendment request:

The amendments would change

Technical Specification (TS) 3/4.6.2,

‘‘Containment Spray System,’’ to verify

on recirculation flow that the

containment spray pumps develop a

differential pressure of at least 204 psi.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change revises the CS

[containment spray] pump technical

specification surveillance test acceptance

from pump discharge pressure to pump

differential pressure. This will account for

the effect of RWST [refueling water storage

tank] level on test results and provide

acceptance criteria that verifies each CS

pump performs as assumed in the accident

analyses. This surveillance test is also being

added to the Salem Unit 1 TS. The proposed

change does not alter the physical plant

arrangement or the method of CS pump

inservice testing. Therefore it does not

increase the probability of an accident. There

is no change to pump performance

requirements as assumed in the accident

analyses. There is no change to CS system

performance in response to an accident.

Therefore, the proposed change does not

involve an increase in the consequences of an

accident previously evaluated.

The proposed change also corrects a

typographical error by removing a repeated

word. This change does not involve an

increase in the consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change revises the Salem

Unit 2 CS pump surveillance test acceptance

criteria from pump discharge pressure to

pump differential pressure. This will account

for the effect of RWST level on test results

and provide acceptance criteria that verify

the CS pumps perform as assumed in the

accident analyses. This surveillance test is

also being added to the Salem Unit 1 TS. The

proposed change does not alter the plant

configuration. The change does not alter the

method of performing inservice testing on the

CS pumps. The change does not alter the CS

pump performance assumed in the accident

analyses. Therefore, the change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change also corrects a

typographical error by removing a repeated

word. This change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change ensures the CS pump

Salem Unit 2 TS surveillance test acceptance

criteria verify CS pump performance as

assumed in the accident analyses accounting

for RWST level effects. This surveillance test

is also being added to the Salem Unit 1 TS.

The proposal does not change the CS pump

performance requirements assumed in the

accident analyses and thus does not reduce

the margin of safety.

The proposed change also corrects a

typographical error by removing a repeated

word. This 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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

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:

November 14, 1997.

Description of amendment request:

The proposed changes to the Technical

Specifications (TSs) include

administrative and editorial changes to

correct errors in the TSs that have either

existed since initial issuance or were

introduced during subsequent changes.

In addition, surveillance requirements

are added that are considered

administrative changes since the

surveillances should have been

incorporated with the TS when the

applicable amendment to the TSs was

approved by the NRC.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to the TS are

administrative or editorial changes to the TS

and do not involve any physical changes to

the plant. The administrative changes and

editorial changes do not delete any existing

surveillance requirements or delete any

requirements from the Limiting Condition for

Operations (LCOs) or Action Statements and

therefore do not reduce the actions that are

currently taken in the TS to demonstrate

operability of plant structures, systems, or

components (SSCs). The additional

surveillance requirements that are being

added to the TS including the new

surveillances correct past administrative

errors and should have been incorporated

within the TS as part of the approved

Amendments to the TS. These changes will

provide additional assurance that SSCs

perform their intended safety functions.

Surveillance testing has been and is currently

being performed for the surveillance

requirements that should have been

incorporated and are now administratively

being added to the TS. Since these changes

do not modify any SSCs or reduce the current

requirements for demonstrating operability of

these SSCs or reduce the current

requirements for demonstrating operability of

these SSCs, the proposed changes to the TS

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.
66142 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

The proposed changes to the TS are

administrative and editorial corrections to

the TS that do not affect the ability of the

plant systems to meet their current TS

requirements or design basis functions. There

is no reduction in the current surveillance

requirements required to demonstrate the

operability of plant SSCs. These changes also

do not involve any physical changes to plant

SSCs. Therefore the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes are administrative

and editorial corrections to the TS that do not

affect the ability of plant SSCs to perform

their design basis accident functions. There

is no reduction in the current surveillance

requirements required to demonstrate the

operability of plant SSCs. 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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit–N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

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:

November 14, 1997.

Description of amendment request:

The proposed changes to the Technical

Specifications (TSs) include

administrative and editorial changes to

correct errors in the TSs that have either

existed since initial issuance or were

introduced during subsequent changes.

In addition, surveillance requirements

are added that are considered

administrative changes since the

surveillances should have been

incorporated with the TS when the

applicable amendment to the TSs was

approved by the NRC.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to the TS are

administrative or editorial changes to the TS

and do not involve any physical changes to

the plant. The administrative changes and

editorial changes do not delete any existing

surveillance requirements or delete any

requirements from the Limiting Condition for

Operations (LCOs) or Action Statements and

therefore do not reduce the actions that are

currently taken in the TS to demonstrate

operability of plant structures, systems, or

components (SSCs). The additional

surveillance requirements that are being

added to the TS including the new

surveillances correct past administrative

errors and should have been incorporated

within the TS as part of the approved

Amendments to the TS. These changes will

provide additional assurance that SSCs

perform their intended safety functions.

Surveillance testing has been and is currently

being performed for the surveillance

requirements that should have been

incorporated and are now administratively

being added to the TS. Since these changes

do not modify any SSCs or reduce the current

requirements for demonstrating operability of

these SSCs or reduce the current

requirements for demonstrating operability of

these SSCs, the proposed changes to the TS

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes to the TS are

administrative and editorial corrections to

the TS that do not affect the ability of the

plant systems to meet their current TS

requirements or design basis functions. There

is no reduction in the current surveillance

requirements required to demonstrate the

operability of plant SSCs. These changes also

do not involve any physical changes to plant

SSCs. Therefore the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes are administrative

and editorial corrections to the TS that do not

affect the ability of plant SSCs to perform

their design basis accident functions. There

is no reduction in the current surveillance

requirements required to demonstrate the

operability of plant SSCs. 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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit–N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket No. 50–311, Salem Nuclear

Generating Station, Unit No. 2, Salem

County, New Jersey

Date of amendment request: October

29, 1997.

Description of amendment request:

The proposed amendment would make

a one-time change to Technical

Specification 3/4.4.6, ‘‘Steam

Generators,’’ to require that the next

inspection be performed within 24

months of criticality for fuel cycle 10,

rather than within 24 months from the

previous inspection. The previous

inspection was performed in May 1996;

thus, adhering to the current Technical

Specification would require inspection

by May 1998 and would require a forced

outage. It would also eliminate

description of an alternate sampling

plan that was applicable only to Unit 2’s

fourth refueling outage.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The Design Basis Accident (DBA) analyzed

in UFSAR Chapter 15.4.4, is Steam Generator

Tube Rupture. The Technical Specification

steam generator tube inspection attempts to

avoid this DBA by maintenance of the

integrity of the primary to secondary coolant

boundary represented by steam generator

tubes. The process by which this integrity is

maintained is inspection of steam generator

tubes at prescribed intervals, and the removal

of defective tubes from service. Inspection

intervals are based on preventing corrosion

growth from exceeding tube structural

strength, thereby preventing tube failure. An

extensive steam generator inspection in May

of 1996 characterized existing steam

generator tube degradation, and degraded

tubes were removed from service at that time.

Degradation growth rates were evaluated for

the next operating interval and it was

determined that full cycle operation would

not challenge tube structural integrity.

Because degraded tubes were plugged, the

integrity of the steam generators has been

restored, and, because further degradation

was prevented by a strictly controlled wet

lay-up program in place since the inspection,

steam generator integrity has since been

maintained at the May 1996 level. This is the

level normally expected for commencement

of full power operations at the beginning of

a fuel cycle. Thus, it can be reasonably
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66143

concluded that this request to extend the

inspection interval to conclude 24 months

after the start of Unit 2 fuel cycle 10 does not

involve an increase in the probability of an

accident previously analyzed.

Salem UFSAR Chapter 15, Section 15.4.4.,

discusses the Design Basis Accident

involving steam generator tube rupture. Since

the Salem Unit 2 steam generators were

extensively inspected and all degraded tubes

were removed from service by plugging,

integrity of the generators was restored to

fully serviceable condition at that time.

Degradation of steam generator tubes has

been prevented since the inspection by a

carefully controlled, EPRI Guidelines based,

corrosion prevention program. It follows,

then, that the Unit 2 steam generators were

in the same condition immediately prior to

fill and vent as if the inspection had just been

concluded. This is the condition assumed for

commencement of normal operation. Thus, it

is reasonable to conclude that this proposal

to extend the current steam generator

inspection interval to end 24 months after

start of Unit 2 fuel cycle 10 represents no

significant increase in the consequences of an

accident previously analyzed.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Steam generator tube inspections

determine tube integrity and provide

reasonable assurance that a tube rupture or

primary to secondary leak will not occur.

Accidents involving steam generator tube

rupture are analyzed in Salem UFSAR

Section 15.4.4, Steam Generator Tube

Rupture. The only type of accident that can

be postulated from extending the steam

generator inspection interval would be a tube

leak or rupture. Thus, it can be concluded

that extending the steam generator inspection

interval on a one-time basis cannot create the

possibility of a 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 margin of safety, as with any TS,

depends upon maintenance of specific

operating parameters within design limits. In

the case of steam generators, that margin is

maintained through assurance of tube

integrity as the primary to secondary

boundary. Assurance of tube integrity is

provided through periodic inservice testing

of tube integrity and removal from service of

defective tubes. Additional margin is

provided through protection from possible

consequences of steam generator tube failure

by detection and mitigation systems. As

discussed in 1., above, there was an extensive

steam generator inspection, and the steam

generators have been maintained since the

inspection, using a lay-up program that

complies with EPRI Guidelines, to prevent

further tube degradation. Also, N–16

monitors were added, enhancing detection

capabilities. The margin as established by the

latest inspection has been maintained by the

corrosion control program of EPRI Primary

and Secondary Guidelines based on wet layup conditions. Thus, it can be reasonably

concluded that this proposal to amend the

Salem Unit 2 Technical Specifications, on a

one-time basis, to extend the steam generator

inspection interval to end 24 months after

start of Unit 2 fuel cycle 10 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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: August 8,

1997.

Description of amendment request:

The proposed amendment would revise

the surveillance requirements (SR) of

Technical Specification (TS) 3/4.7.4

‘‘Essential Service Water System’’ by

removing the requirement to perform SR

4.7.4.b.1, 4.7.4.b.2 and 4.7.4.c during

shutdown.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change to TS has no adverse

impact on the probability of occurrence or

the consequences of an accident. The

proposed amendment does not change or

alter the design assumptions for the systems

or components used to mitigate the

consequences of an accident and the

methodologies used in the accident analysis

remain unchanged. The operating limits and

the radiological consequences will not be

changed. No design basis accidents will be

affected by this change since the required TS

surveillances will continue to be performed

on an 18 month frequency.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

All design and performance criteria

continue to be met and no new failure

mechanisms have been identified. The

proposed change does not affect the design

or operation of any system or component in

the plant since the required TS surveillances

will continue to be performed on an 18

month frequency. The safety functions of the

related structures, systems or components are

not changed in any manner, nor is the

reliability of any structure, system or

component reduced. Conducting these

surveillances online will not increase the

possibility of plant transients. Since the

safety functions and reliability are not

adversely affected, the proposed change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change will not affect or

change a safety limit or affect plant

operations since the required TS

surveillances will continue to be performed

on an 18 month frequency. This change will

not reduce the margin of safety assumed in

the accident analysis nor reduce any margin

of safety as defined in the basis for any TS.

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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: William H.

Bateman.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: August 8,

1997.

Description of amendment request:

The proposed amendment would revise

Table 3.3–3, Functional Units 4.b.2 and

5.a.2 of the Callaway Technical

Specifications (TS) by (1) changing the

main steam and feedwater isolation

system (MSFIS) channels to be

consistent with the requirements for the

solid state protection system (SSPS), (2)

adding a clarifying note, and (3) deleting

and replacing Action Statements 27a

and 34a with Action Statements 27 and

34. In addition, Table 4.3–2, Functional

Units 4.b and 5.a are proposed to be

revised by changing the slave relay

quarterly surveillance to a quarterly

actuation logic test for the MSFIS

actuation and relays.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
66144 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to Technical

Specifications (TS) have no adverse impact

on the probability of occurrence or the

consequences of an accident. The proposed

amendment does not change or alter the

design assumptions for the systems or

components used to mitigate the

consequences of an accident and the

methodologies used in the accident analysis

remain unchanged. The operating limits and

the radiological consequences will not be

changed. No design basis accidents will be

affected by these changes. The proposed

changes do not result in any hardware

changes.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. All design and performance

criteria continue to be met and no new

failure mechanisms have been identified. The

proposed changes do not affect the design or

operation of any system or component in the

plant. The safety functions of the related

structures, systems or components are not

changed in any manner, nor is the reliability

of any structure, system or component

reduced. However, these changes are

consistent with the requirements for the

SSPS. Since the safety functions and

reliability are not adversely affected, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes will not affect or

change a safety limit or affect plant

operations. These changes will not reduce

the margin of safety assumed in the accident

analysis nor reduce any margin of safety as

defined in the basis for any TS. The proposed

changes do not affect the acceptance criteria

for any analyzed event. No setpoints are

revised and the system response time will

not be affected.

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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: William H.

Bateman.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: August 8,

1997.

Description of amendment request:

The proposed amendment would revise

Table 3.7–2 of the Technical

Specifications to specify that the lift

setting tolerance for the main steam line

safety valves be +3/¥1% as-found and

plus or minus 1% as-left. Table 2.2–1

would be revised by reducing the sensor

error for the pressurizer pressure-high

trip.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The main steam line safety valves are

designed to mitigate transients by preventing

overpressurization of the main steam system.

The proposed change does not alter this

design basis. The revised analysis shows that

the probability or consequences of all

previously analyzed accidents are not

changed by increasing the setpoint tolerance

of the safety valves. Therefore, there is no

increase in the probability of occurrence or

the consequences of any accident.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There is no new type of accident or

malfunction created, the method and manner

of plant operation will not change nor is

there a change in the method in which any

safety related system performs its function.

Any main steam safety valve lifting at the

extremes of the proposed tolerance will not

result in a low lift setpoint that is less than

the normal no load system pressure or a high

lift setpoint that allows main steam system

overpressurization.

3. The proposed change does not involve

a significant reduction in a margin of safety.

This is based on the fact that no plant

design changes are involved and the method

and manner of plant operation remains the

same. With the increased setpoint tolerance,

the main steam safety valves will still

prevent pressure from exceeding 110 percent

of design pressure in accordance with the

ASME code. All FSAR accident analysis

conclusions remain valid and unaffected by

this change.

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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037

NRC Project Director: William H.

Bateman.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: August 8,

1997.

Description of amendment request:

The proposed amendment application

would revise feedwater isolation

engineered safety feature actuation

system (ESFAS) functions in Technical

Specification Tables 3.3–3, 3.3–4 and

4.3–2 as follows:

(1) The Applicable MODES for

Functional Units 5.a.1), Automatic

Actuation Logic and Actuation Relays,

and 5.a.2), Automatic Actuation Logic

and Actuation Relays, in Tables 3.3–3

and 4.3–2 would be revised to add

MODE 3.

(2) A new Functional Unit 5.d, Steam

Generator (SG) Water Level Low-Low

(for feedwater isolation only), would be

added to Tables 3.3–3, 3.3–4, and 4.3–

2.

(3) In conjunction with the changes

under item (2), the Applicable MODES

in Table 3.3–3 for AFW SG Water Level

Low-Low Functional Units 6.d.1).c),

Start Motor-Driven Pumps Vessel delta

T (Power-1, Power-2), would be revised

to delete MODE 3. Functional Unit

6.d.3) in Table 4.3–2 would also be

revised to delete MODE 3.

(4) The Bases for Functional Unit

11.b, Reactor Trip P–4, in Table 3.3–3

would be revised to add a note allowing

the feedwater isolation function on P–4

coincident with low Tavg to be blocked.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Actuation Logic Applicability and New SG

Water Level Low-Low Functional Unit

1. The proposed change does not involve

a significant increase in the probability or
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66145

consequences of an accident previously

evaluated.

The proposed changes impose more

stringent requirements and have been

reviewed to ensure no previously evaluated

accident has been adversely affected. The

more stringent requirements are imposed to

ensure the plant’s operation and testing are

consistent with the safety analysis and

licensing basis. Therefore, the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not involve a

physical alteration of the plant (no new or

different type of equipment will be installed

other than the bypass switch addressed in a

separate 50.92 evaluation below) or changes

in controlling parameters. The proposed

changes do impose different requirements;

however, these changes are consistent with

assumptions made in the safety analysis and

licensing basis. Actuation logic applicability

is extended to MODE 3 and the SSPS slave

relays that implement feedwater isolation on

SG water level low-low will continue to be

surveilled quarterly as they have always been

tested. Thus, the proposed changes do 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 imposition of more stringent

requirements does not reduce the margin of

safety. The margin of safety would be

increased since the scope of the Technical

Specifications has been increased to include

additional plant equipment and add

additional Applicability requirements. The

changes are consistent with the safety

analysis and licensing basis. Therefore, the

proposed changes do not involve a reduction

in a margin of safety.

TTD Applicability

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Overall protection system performance will

remain within the bounds of the previously

performed accident analyses since no

hardware changes are proposed. The

proposed change adds a relaxation to the

Applicability for the SG Water Level LowLow Vessel delta T channels. The proposed

change in the Applicability will not affect

any of the analysis assumptions for any of the

accidents previously evaluated. The

proposed change will not affect the

probability of any event initiators nor will

the proposed change affect the ability of any

safety-related equipment to perform its

intended function. A Vessel delta T channel

should only be tripped if it is inoperable and

the reactor is operating, when the need to

restrict trip time delays is applicable. There

will be no degradation in the performance of

nor an increase in the number of challenges

imposed on safety-related equipment

assumed to function during an accident

situation. Accident analyses have been

performed with the maximum trip time

delays enabled at power levels up to 19%

RTP (10% RTP plus uncertainty). Therefore,

operation in MODE 3 with the maximum trip

time delays is enveloped. Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There are no hardware changes nor are

there any changes in the method by which

any safety-related plant system performs its

safety function. The change in Applicability

will not impact the normal method of plant

operation. The maximum trip time delay

should be enabled in MODE 3 to preclude an

unnecessary feedwater isolation or auxiliary

feedwater actuation from occurring prior to

the expiration of the trip time delay

previously analyzed for MODE 1 operation.

No new accident scenarios, transient

precursors, failure mechanisms, or limiting

single failures are introduced as a result of

this change. Therefore, the proposed change

does not create the possibility of a new of

different kind of accident from any

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not affect the

acceptance criteria for any analyzed event.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems necessary to

assure the accomplishment of protection

functions. There will be no impact on any

margin of safety.

Feedwater Isolation on P–4/Low Tavg Bypass

Switch

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Overall protection system performance will

remain within the bounds of the previously

performed accident analyses. The P–4/Low

Tavg Bypass Switch design change will not

impact any accidents previously evaluated in

the FSAR since feedwater isolation upon

reaching this function was never credited.

The ESFAS will continue to function in a

manner consistent with the accident analysis

assumptions and the plant design basis. As

such, there will be no degradation in the

performance of nor an increase in the number

of challenges to equipment assumed to

function during an accident situation.

This Technical Specification change does

not affect the probability of any event

initiators. There will be no change to normal

plant operating parameters or accident

mitigation capabilities. Therefore, there will

be no increase in the probability or

consequences of any accident occurring due

to this change.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There are no changes in the method by

which any safety-related plant system

performs its safety function and the normal

manner of plant operation is unaffected,

other than the proposed allowance to bypass

feedwater isolation on P–4 coincident with

low Tavg. This bypass switch modification

will be performed under the design standards

applicable to all safety system bypasses at

Callaway, except for Section 4.12 of IEEE

279–1971. Section 4.12 of IEEE 279–1971

requires that an operating bypass of a

protective function be automatically removed

whenever permissive conditions are not met.

However, the subject circuitry does not

provide a protective function. It is not

assumed or credited in any safety analysis. In

addition, plant conditions that would call for

the restoration of the feedwater isolation

function cannot occur without operator

action to close the reactor trip breakers.

Administrative controls will govern the

proper use of and restoration from the

proposed bypass. Although the addition of

the bypass switch introduces the potential for

an equipment malfunction of a different type

from any previously evaluated in the FSAR,

the possibility of a new or different type of

accident is not created. The switch functions

only to allow a manual bypass of feedwater

isolation. The failure of the switch or its

improper use will not be an event initiator

for the previously analyzed Loss of Normal

Feedwater event in FSAR Section 15.2.7

since it cannot fail in such a manner as to

cause feedwater isolation.

No new accident scenarios, transient

precursors, failure mechanisms, or limiting

single failures are introduced as a result of

this change. There will be no adverse effect

or challenges imposed on any safety-related

system as a result of this change. Therefore,

the possibility of a new or different type of

accident is not created.

3. The proposed change does not involve

a significant reduction in a margin of safety.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems necessary to

assure the accomplishment of protection

functions. There will be no impact on DNBR

limits, FQ, F-delta-H, LOCA PCT, peak local

power density, or any other 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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: William H.

Bateman.
66146 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

Virginia Electric and Power Company,

Docket Nos. 50–338 and 50–339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

November 5, 1997.

Description of amendment request:

The current Technical Specifications

requirements prohibit loads in excess of

2500 pounds from traveling over

irradiated fuel assemblies in the spent

fuel pit. Due to the number of irradiated

fuel assemblies currently stored in the

spent fuel pit over years of operation,

additional flexibility is needed to

accomplish the movement of the spent

fuel pit gates during refueling activities

and to reduce fuel handling activities in

preparation for refueling outages. In

order to perform gate seal maintenance

prior to each outage, a gate is moved

across the irradiated fuel storage area to

the cask handling area where it can be

lifted out of the spent fuel pit. When a

clear path of empty fuel storage cells

cannot be established, seal maintenance

cannot be performed unless relief from

the current Limiting Condition of

Operation is granted. The proposed

changes will exempt these requirements

for the movements of the spent fuel

gates provided specific administrative

controls are satisfied.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Specifically, operation of the North Anna

Power Station in accordance with the

proposed changes will not:

1. Involve a significant increase in the

probability of occurrence or consequences of

an accident previously evaluated.

The accident in question is a fuel handling

accident in the spent fuel pit. The proposed

changes will actually reduce the probability

of a fuel handling accident by eliminating

unnecessary fuel assembly movements. After

this change is implemented, only those

assemblies containing control rod assemblies

will be subjected to such moves prior to

movement of the gates instead of the current

practice of moving all the fuel necessary to

establish a load path of empty cells. A

redundant rigging system will be provided

which eliminates the possibility of a load

drop due to a hoist failure. Furthermore, even

though the double rigging system makes a

load drop due to a hoist failure an incredible

event, a calculation was performed to

determine the effects of a direct impact load

on a single fuel storage cell or the SFP [spent

fuel pit] structure. The calculation concludes

that there will be no adverse consequences to

either irradiated fuel or the SFP structure.

The plant design basis fuel handling accident

will not be violated. Therefore, with the

administrative controls in place to eliminate

the possibility of a gate drop the probability

of occurrence or the consequences of a fuel

handling accident are not increased.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes establish adequate

administrative controls over the spent fuel

pit gate movements to prevent damage to

stored irradiated fuel and fuel racks thereby

ensuring the design basis fuel handling

accident remains bounding and that fuel

spacing is maintained in the racks precluding

criticality.

3. Involve a significant reduction in any

margin of safety.

The new administrative controls ensure

that a postulated gate drop will not occur due

to compliance with our licensing

commitments to NUREG–0612 and the

requirement to install a redundant rigging

system to eliminate the possibility of a load

drop initiated by hoist failure. Analysis has

determined that in the event the gate was to

be dropped from its controlled lift height: (1)

There will be no damage to irradiated fuel

caused by the direct impact loading on a

single storage cell and (2) the fuel storage

rack will maintain fuel in a non-critical array.

A new criteria, demonstrating the ability of

the pool floor to remain intact after a gate

drop has been shown by analysis. New

controls prevent the degradation of the

existing margin of safety and ensure an

adequate safety margin for the new criteria.

The administrative controls added for the

gate lift preclude the possibility of a load

drop induced by a hoist failure and, therefore

ensure the potential for radioactivity release

and inadvertent criticality remain bounded

by the present design basis. Therefore, the

margin of safety is not reduced by the

proposed change.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: James E. Lyons.

Virginia Electric and Power Company,

Docket Nos. 50–338 and 50–339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

November 18, 1997.

Description of amendment request:

The Technical Specifications

surveillance requirements currently

require testing and inspection of the

Turbine Overspeed Protection System

control valves, at least once per 31 days,

to ensure their ability to prevent

overspeeding of the turbine. Based on

an analysis of Westinghouse BB–296

turbines with steam chests, the

proposed change would increase the

surveillance test interval from at least

once per 31 days to at least once per 92

days.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Specifically, operation of the North Anna

Power Station in accordance with the

proposed Technical Specifications changes

will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

No new or unique accident precursors are

introduced by these changes in surveillance

requirements. The probability of turbine

missile ejection with an extended test

interval to 92 days for the turbine governor

and throttle valves has been determined to

remain within the applicable NRC

acceptance criteria. The heavy hub design of

the turbine rotors provides further assurance

that the probability of ejection of turbine

missiles due to destructive overspeed

remains within the acceptance criteria.

Therefore, these changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The demonstrated high reliability of the

turbine governor and throttle valves and the

verification of the operability of the other

turbine control valves provide adequate

assurance that the turbine overspeed

protection system will operate as designed, if

needed. Turbine governor and throttle valve

testing performed to date has demonstrated

the reliability of these valves. In addition, the

operability of the other turbine valves (i.e.,

reheat and intercept stop valves) will

continue to be verified every 18 months as

required by the Technical Specifications.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Since the implementation of the proposed

change to the surveillance requirements will

not require hardware modifications (i.e.,

alterations to plant configuration), operation

of the facilities with these proposed

Technical Specifications does not create the

possibility for any new or different kind of

accident which has not been already been

evaluated in the Updated Final Safety

Analysis Report (UFSAR). In addition, the

results of the probabilistic evaluation

indicate that no additional transients have

been introduced.

The proposed revision to the Technical

Specifications will not result in any physical

alteration to any plant system, nor would

there be a change in the method by which

any safety-related system performs its

function. The design and operation of the
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66147

turbine overspeed protection and turbine

control systems are not being changed.

The proposed Technical Specifications

changes do not affect the design, operation,

or failure modes of the valves and other

components of the turbine overspeed

protection system. Therefore, the proposed

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes do not reduce the

margin of safety as defined in the basis for

any Technical Specifications. Furthermore,

the total turbine missile ejection probability

continues to be enveloped by the applicable

acceptance criteria of 1E–5. The design and

operation of the turbine overspeed protection

and turbine control systems are not being

changed and the operability of the turbine

governor and throttle valves will be

demonstrated on a refuelling outage basis. In

addition, the results of the accident analyses,

which are documented in the UFSAR,

continue to bound operation with the

proposed change in surveillance interval for

the turbine throttle and governor valves, so

that there is no safety margin reduction.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: James E. Lyons.

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.

Tennessee Valley Authority, Docket

Nos. 50–327, Sequoyah Nuclear Plant,

Unit 1, Hamilton County, Tennessee

Date of application for amendments:

November 21, 1997.

Description of amendments request:

Amend Technical Specifications to add

a one-time allowance through Operating

Cycle 9 to Surveillance Requirement

4.4.3.2.1.b to perform stroke testing of

the power-operated relief valve in Mode

5 rather than in Mode 4.

Date of publication of individual

notice in the Federal Register:

December 1, 1997 (62 FR 63565).

Expiration date of individual notice:

December 31, 1997.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

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.

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

November 30, 1995, as supplemented

March 15, 1996, March 6, 1997, and

June 27, 1997.

Brief description of amendments: The

amendments incorporate references to a

new Combustion Engineering, Inc.

topical report describing steam

generator tube sleeves, delete references

to the previous CE topical report,

incorporate sleeve/tube inspection

scope and expansion criterion, revise

the plugging limit for a CE sleeve to

28% of the nominal sleeve wall

thickness, and incorporate a post weld

heat treatment for free span welds.

Date of issuance: November 18, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 223 and 199.

Facility Operating License Nos. DPR–

53 and DPR–69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 3, 1996 (61 FR 176).

The March 15, 1996, March 6, 1997, and

June 27, 1997, letters provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated November 18,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et

al., Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of application for amendments:

August 6, 1997.

Brief description of amendments: The

amendments address an unreviewed

safety question associated with the

handling of the spent fuel shipping cask

at the Brunswick Steam Electric Plant,

Units 1 and 2.

Date of issuance: December 2, 1997.

Effective date: December 2, 1997.
66148 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

Amendment Nos.: 190 and 221.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments authorize

changes to the facility’s Updated Final

Safety Analysis Report.

Date of initial notice in Federal

Register: September 17, 1997 (62 FR

48897) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 2, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

February 18, 1997.

Brief description of amendment: This

amendment revises the maximum

allowable power range neutron flux

high setpoints (percent of rated thermal

power) shown in Technical

Specification Table 3.7–1.

Date of issuance: November 25, 1997.

Effective date: November 25, 1997.

Amendment No.: 75.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17225)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 25, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Date of application for amendments:

April 7, 1997, as supplemented on

August 7, 1997.

Brief description of amendments: The

amendments revise the technical

specifications to permit installation and

use of C&D Charter Power Systems, Inc.,

batteries.

Date of issuance: November 25, 1997.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 93 and 93.

Facility Operating License Nos. NPF–

37 and NPF–66: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54868). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

November 25, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Byron Public Library District,

109 N. Franklin, P.O. Box 434, Byron,

Illinois 61010.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

October 13, 1997, as supplemented by

letters dated October 28 and November

5, 1997.

Brief description of amendments: The

amendments revise TS Table 3.3–4,

‘‘Engineered Safety Features [ESF]

Actuation System Instrument Trip

Setpoints.’’ Specifically, the

amendments support the replacement of

three safety-related narrow range

Refueling Water Storage Tank level

instruments with three safety-related

wide range level instruments. The ESF

trip setpoint for the refueling water

automatic switchover to recirculation is

revised to account for the difference in

instrument uncertainty associated with

wide range level instruments and

provides additional operator response

time margin.

Date of issuance: November 25, 1997.

Effective date: Unit 1—As of the date

of issuance to be implemented

consistent with the refueling outage

scheduled for June 1998; Unit 2—As of

the date of issuance to be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 177 (Unit 1); 159

(Unit 2).

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54859). The October 28 and November

5, 1997, letters provided additional and

clarifying information that did not

change the scope of the October 13,

1997, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 25,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application for amendments:

October 10, 1997, as supplemented by

letters dated November 3, 6, and 10,

1997.

Brief description of amendments: The

amendments revise Technical

Specifications to implement alternate

repair criteria for steam generator tubes

that have degraded roll joints inside of

the upper tubesheet. The alternate

repair criteria would allow new roll

joints to be installed below the degraded

roll joints in the upper tubesheet.

Date of issuance: November 21, 1997.

Effective date: November 21, 1997.

Amendment Nos.: Unit 1—227; Unit

2—227; Unit 3—224.

Facility Operating License Nos. DPR–

38, DPR–47, AND DPR–55: The

amendments revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes. (62 FR 55835 dated

October 28, 1997). The notice provided

an opportunity to submit comments on

the Commission’s proposed no

significant hazards consideration

determination. No comments have been

received. The notice also provided for

an opportunity to request a hearing by

November 28, 1997, but indicated that

if the Commission makes a final no

significant hazards consideration

determination, any such hearing would

take place after issuance of the

amendments.

The Commission’s related evaluation

of the amendments, finding of exigent

circumstances, and a final

determination of no significant hazards

consideration are contained in a Safety

Evaluation dated November 21, 1997.

Attorney for licensee: M. J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina.

Duquesne Light Company, et al., Docket

No. 50–334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of application for amendment:

March 10, 1997, as supplemented July

28 and September 17, 1997.

Brief description of amendment: The

amendment modifies Technical

Specification 3/4.4.5, ‘‘Steam

Generators,’’ and its associated Bases
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66149

and adds a new license condition to

Appendix C for Beaver Valley Power

Station, Unit No. 1 (BVPS–1) to allow

repair of steam generator tubes by

installation of sleeves developed by

ABB Combustion Engineering. In

addition, the amendment deletes the

option for using the kinetic sleeving

methodology previously approved for

use at BVPS–1.

Date of issuance: November 25, 1997.

Effective date: As of date of issuance,

to be implemented within 60 days.

Amendment No.: 208.

Facility Operating License No. DPR–

66: Amendment revised the Technical

Specifications and License.

Date of initial notice in Federal

Register: April 23, 1997 (62 FR 19829).

The July 28 and September 17, 1997,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination or expand the

amendment request beyond the scope of

the April 23, 1997, Federal Register

notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 25,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

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 12, 1997, as supplemeneted

November 7, 1997.

Brief description of amendment: The

proposed amendment involves a

revision to the Emergency Diesel

Generator protective relaying scheme at

Crystal River Unit 3, to be reflected in

the next revision to the Final Safety

Analysis Report (FSAR).

Date of issuance: December 1, 1997.

Effective date: Effective upon

issuance.

Amendment No.: 159.

Facility Operating License No. DPR–

72:. Amendment revises the FSAR.

Date of initial notice in Federal

Register: September 30, 1997 (62 FR

51165). By letter dated November 7,

1997, the licensee provided additional

information which did not affect the

original no significant hazards

consideration.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 1,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

October 10, 1996, as supplemented

March 25, June 6, and August 29, 1997.

Brief description of amendment: The

amendment extends the instrumentation

surveillances for the condenser low

vacuum, high temperature main

steamline tunnel, recirculation flow,

and reactor coolant leakage.

Additionally, the change extends the

equipment test/operability checks for

containment vent and purge isolation,

electromagnetic relief valve operability,

and drywell to torus leakage test.

Date of Issuance: November 26, 1997.

Effective date: November 26, 1997,

with full implementation within 60

days.

Amendment No.: 193.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57485). The Commission’s related

evaluation of this amendment is

contained in a Safety Evaluation dated

November 26, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit 2, New

London County, Connecticut

Date of application for amendment:

May 20, 1997, as supplemented on

September 23, 1997.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by relocating the

containment isolation valve (CIV) list

from the TSs to the Technical

Requirements Manual in accordance

with Generic Letter 91–08, ‘‘Removal of

Component Lists from the Technical

Specifications.’’ The amendment also

changes the surveillance requirement

for valves, blind flanges, and

deactivated automatic valves located

inside containment that are locked,

sealed, or otherwise secured in the

closed position from once every 31 days

to during each cold shutdown, but no

more than once per 92 days. The TS

Bases is changed to reflect the relocation

of the containment isolation valve list

from the TSs to the Technical

Requirements Manual and dicusses

administrative controls for CIV

operation in Modes 1 through 4. Also,

a license condition has been added to

paragraph 2.C. of the Operating License

to ensure enforceability and to provide

a method of tracking the license

condition back to the license

amendment.

Date of issuance: November 19, 1997.

Effective date: As of the date of

issuance, to be implemented within 90

days.

Amendment No: 210

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications and License Conditions.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33128).

The September 23, 1997, letter provided

clarification of the initial submittal and

did not affect the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

November 19, 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, Connecticut and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut.

Date of application for amendment:

September 16, 1997.

Brief description of amendment: The

amendment changes the main steam

line American Society of Mechanical

Engineers Code (Code) safety valves

Technical Specifications (TSs) by: (1)

Deleting TS Table 3.7.1, ‘‘Maximum

Allowable Power Level-High Trip

Setpoint with Inoperable Steam Line

Safety Valves During Operation with

Both Steam Generators,’’ by not

allowing operation in Mode 1 or 2 with

inoperable Code safety valves while

allowing operation in Mode 3 with up

to three Code safety valves inoperable

per steam generator, (2) modifing the

associated action statement in TS 3.7.1.1

to reflect the operational changes, and

(3) updating the TS Bases to reflect the

proposed changes and include the

correct amendment history numbers to
66150 Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices

reflect previously approved

amendments.

Date of issuance: November 19, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 211.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 8, 1997 (62 FR 52582).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 19, 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, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

September 26, 1997.

Brief description of amendments: The

amendments revise Technical

Specification (TS) 3.4.B, ‘‘Auxiliary

Feedwater System,’’ to provide specific

guidance for conducting postmaintenance operational testing of the

turbine-driven auxiliary feedwater

pump and associated system valves to

meet operability requirements and

limiting conditions for operation during

unit startup. Additionally, the

amendments revise Table TS.3.5.2B to

allow the auxiliary feedwater pump

auto-start actuation instrumentation to

be bypassed during startup and

shutdown operations when the main

feedwater pumps are not required to

supply feedwater to the steam

generators.

Date of issuance: November 25, 1997.

Effective date: November 25, 1997,

with full implementation within 30

days.

Amendment Nos.: 134 and 126.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54874). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

November 25, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 20,

1996.

Brief description of amendment: The

amendment revises the technical

specifications to correct and clarify

surveillance test requirements for the

reactor protective system and other

plant instrumentation and control

systems.

Date of issuance: November 24, 1997.

Effective date: November 24, 1997, to

be implemented within 120 days of the

date of issuance.

Amendment No.: 182.

Facility Operating License No. DPR–

40: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 28, 1996 (61 FR

44361). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 24, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of application for amendment:

May 10, 1996, as supplemented March

19 and August 29, 1997.

Brief description of amendment: The

amendment incorporates into the

Technical Specifications the Margin

Recovery portion of the Fuel Upgrade

Margin Recovery Program and support

increased steam generator plugging,

improved fuel reliability, reduced fuel

costs, longer fuel cycles, reduced spent

fuel pool storage, and enhanced reactor

safety.

Date of issuance: November 26, 1997.

Effective date: As of date of issuance.

To be implemented on Unit 1 prior to

entry into Mode 2 from the current

outage.

Amendment No.: 201.

Facility Operating License No. DPR–

70: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34898).

The March 19 and August 29, 1997,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

November 26, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Southern California Edison Company,

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

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

January 4, 1996.

Brief description of amendments:

These amendments delete License

Condition 2.C(26) for SONGS Unit 2 and

License Condition 2.C(27) for SONGS 3.

These license conditions require that

Southern California Edison implement

and maintain a plan for scheduling all

capital modifications based on an NRC

approved Integrated Implementation

Schedule Program Plan.

Date of issuance: December 3, 1997.

Effective date: December 3, 1997.

Amendment Nos.: Unit 2—137; Unit

3—129.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Facility Operating Licenses.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR 15997).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated December 3, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

January 20, 1997.

Brief description of amendment: This

amendment revises Technical

Specification (TS) Section 3/4.5.2,

‘‘Emergency Core Cooling Systems,

ECCS Subsystems¥Tavg greater than or

equal to 280°F,’’ TS Section 3/4.5.3,

‘‘Emergency Core Cooling Systems,

ECCS Subsystems¥Tavg less than 280

°F,’’ and TS Section 3/4.7, ‘‘Plant

Systems.’’ Several surveillance intervals

were changed from 18 months to once

each refueling interval.

Date of issuance: December 2, 1997.
Federal Register / Vol. 62, No. 242 / Wednesday, December 17, 1997 / Notices 66151

Effective date: December 2, 1997.

Amendment No.: 216

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 12, 1997 (62 FR 11498).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 2, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

September 17, 1996, as supplemented

by letters dated November 27, 1996, and

October 14, 1997.

Brief description of amendment: This

amendment revises the surveillance

interval from 18 months to less than or

equal to 730 days, nominally 24 months,

for Technical Specification (TS) 3/4.5.2,

‘‘Emergency Core Cooling Systems—

ECCS Subsystems—Tavg greater than or

equal to 280 degrees F’’; TS 3/4.6.5.1,

‘‘Containment Systems—Shield

Building—Emergency Ventilation

System’’; TS 3/4.7.6.1, ‘‘Plant Systems—

Control Room Emergency Ventilation

System’’; TS 3/4.7.7, ‘‘Plant Systems—

Snubbers’’; TS 3/4.9.12, ‘‘Refueling

Operations—Storage Pool Ventilation’’;

and TS Bases 3/4.7.7—‘‘Snubbers.’’

Date of issuance: December 2, 1997.

Effective date: Immediately, and shall

be implemented no later than 120 days

after issuance.

Amendment No.: 217.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 9, 1996 (61 FR 52972).

The supplemental information

submitted by the licensees did not

impact the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 2,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

December 11, 1996 (as supplemented by

letter dated January 6, 1997), January 30,

1997 (as supplemented by letter dated

September 15, 1997), and April 18,

1997.

Brief description of amendment: This

amendment extends surveillance

requirement intervals from 18 to 24

months, revises setpoints, and revises

TS 2.2, ‘‘Limiting Safety System

Settings.’’ Administrative changes have

also been made.

Date of issuance: December 2, 1997.

Effective date: December 2, 1997.

Amendment No.: 218.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Dates of initial notice in Federal

Register: January 15, 1997 (62 FR 2194),

March 12, 1997 (62 FR 11498) and June

4, 1997 (62 FR 30654). The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 2, 1997.

No significant hazards consideration

comments received: No. The

supplemental information provided by

the licensees did not affect the proposed

no significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606.

Dated at Rockville, Maryland, this 10th day

of December 1997.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 97–32763 Filed 12–16–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–315 and 50–316]

American Electric Power Company;

Receipt of Petition for Director’s

Decision Under 10 CFR 2.206

Notice is hereby given that by Petition

dated October 9, 1997, David A.

Lochbaum, on behalf of the Union of

Concerned Scientists, has requested that

the U.S. Nuclear Regulatory

Commission (NRC) take action with

regard to Donald C. Cook Nuclear Plant,

Units 1 and 2, operated by American

Electric Power Company (the Licensee).

The Petition requests that the operating

licenses for D.C. Cook, Units 1 and 2, be

modified, revoked, or suspended until

there is reasonable assurance that the

Licensee’s systems are in conformance

with design-and licensing-bases

requirements. The Petition requests that

systems with a safety function at D.C.

Cook be qualified and capable of

performing their required function

under all design conditions before

restart. The Petition also requests that a

public hearing into this matter be held

in the Washington, DC, area before the

first unit at D.C. Cook is authorized to

restart.

As the basis for these requests, the

Petition states that the NRC recently

completed an architect/engineer design

inspection at D.C. Cook. The Licensee

had previously reviewed the same

systems as part of its design-basis

documentation reconstitution program.

Findings by the NRC inspection team

led to a shutdown of both units and has

necessitated changes to the plant’s

physical configuration. Therefore, the

Petition asserts that the Licensee’s

design-basis documentation

reconstitution and updated final safety

analysis report validation programs lack

the necessary rigor and focus. The

Petition further asserts that deficiencies

in the Licensee’s design control

programs may also be responsible for

similar problems in its safety systems,

which were not examined by the NRC.

The request is being treated pursuant

to 10 CFR 2.206 of the Commission’s

regulations. The request has been

referred to the Director of the Office of

Nuclear Reactor Regulation. As

provided by 10 CFR 2.206, appropriate

action will be taken on this Petition

within a reasonable time. A copy of the

Petition is available for public

inspection at the Commission’s Public

Document Room, located at the Gelman

Building, 2120 L Street, NW.,

Washington, DC 20555–0001.

Dated at Rockville, Maryland, this 9th day

of December 1997.

For The Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–32878 Filed 12–16–97; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Submission for OMB Review;

Comment Request

Upon Written Request, Copies Available

From: Securities and Exchange