Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30629

(Public Meeting) (if needed)

2:00 p.m.—Briefing on Salem (Public

Meeting), (Contact: John Zwolinski,

301–415–1453)

The schedule for Commission

meetings is subject to change on short

notice. To verify the status of meetings

call (recording)—(301) 415–1292.

Contact person for more information:

Bill Hill (301) 415–1661.

* * * * *

The NRC Commission Meeting Schedule

can be found on the Internet at: http://

www.nrc.gov/SECY/smj/schedule.htm.

This notice is distributed by mail to

several hundred subscribers: if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary. Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661).

In addition, distribution of this

meeting notice over the internet system

is available. If you are interested in

receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

* * * * *

Dated: May 30, 1997.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 97–14679 Filed 6–2–97; 10:21 am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from May 12,

1997, through May 22, 1997. The last

biweekly notice was published on May

21, 1997 (62 FR 27792).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By July 7, 1997, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first
30630 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Consumers Power Company, Docket

No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request: April 22,

1997 (supersedes October 15, 1996,

request)

Description of amendment request:

The proposed amendment would revise

the Big Rock Point Technical

Specifications to correct several

administrative and editorial

inconsistencies.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes are clarifications

within the Technical Specifications, and do

not alter the technical content of the

technical specifications. Plant operation or

configuration is not affected. The postulated

doses received by the general public and

plant personnel as a direct result of accidents

previously described, are not affected. Plant

operation or configuration is not affected.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes are either

clarifications to correct inconsistencies

within the Technical Specifications, or

corrections of typographical errors. The

proposed changes do not alter the facility in

any way, therefore the proposed changes do

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 change[s] [do] not affect any

margin of safety as defined by the Plant

Technical Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: John N. Hannon

Consumers Power Company, Docket

No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request: April 30,

1997

Description of amendment request:

The proposed amendment would alter

the company name in the Facility

Operating License DPR-6 and Technical

Specifications for the Big Rock Point

Plant. Specifically, the proposed

amendment would revise the company

name from ‘‘Consumers Power

Company’’ to ‘‘Consumers Energy

Company.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30631

The proposed changes alter the company

name in the Facility Operating License and

Technical Specifications to reflect the change

from ‘‘Consumers Power Company’’ to

‘‘Consumers Energy Company’’. The

company will continue to own all of the

same assets, will continue to serve the same

customers, and will continue to honor all

existing obligations and commitments.

Since the proposed changes do not alter

the technical content of any Facility

Operating License or Technical

Specifications requirements, they do not alter

the design, function, or operation of any

plant structure, system, or component.

Therefore, the changes will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes alter the company

name in the Facility Operating License and

Technical Specifications to reflect the change

from ‘‘Consumers Power Company’’ to

‘‘Consumers Energy Company’’. The

company will continue to own all of the

same assets, will continue to serve the same

customers, and will continue to honor all

existing obligations and commitments.

Since the proposed changes do not alter

the technical content of any Facility

Operating License or Technical

Specifications requirements, they do not alter

the design, function, or operation of any

plant structure, system, or component.

Therefore, the changes will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in a

margin of safety.

Since the proposed changes do not alter

the technical content of any Facility

Operating License or Technical

Specifications requirements, they do not alter

the design, function, or operation of any

plant structure, system, or component.

Therefore, the changes will not involve a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: John N. Hannon

Duke Power Company, Docket No. 50-

413, Catawba Nuclear Station, Unit 1,

York County, South Carolina

Date of amendment request: May 8,

1997

Description of amendment request:

The proposed amendment would add a

phrase to the footnote to Section 3.4.1.2

of the Technical Specifications that

would permit all reactor coolant pumps

(RCPs) to be deenergized for up to 4

hours during Mode 3 on a one-time

basis. Currently, the RCPs are permitted

to be deenergized for up to 1 hour

during Mode 3. The proposed change

would allow the licensee to perform a

natural circulation test using the new

steam generators (installed in late 1996).

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 activity does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed natural circulation test

would be performed in Mode 3 with the

reactor subcritical. This transient is bounded

by the transient analyzed in UFSAR

[Updated Final Safety Analysis Report]

Section 15.2.6, Loss of Non-Emergency AC

Power to the Station Auxiliaries. For this

ANS [American Nuclear Society] Condition II

event, the reactor is assumed to be operating

at 102% power, the turbine driven auxiliary

feedwater pump is assumed unavailable and

each steam generator is assumed to have 18%

of the steam generator tubes plugged. By

contrast, the planned natural circulation test

would be performed with the reactor

subcritical, less than 0.1% of the tubes

plugged in each steam generator, and all

support systems such as auxiliary feedwater,

operable for the test. Therefore, the proposed

natural circulation test would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2) The activity does not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

The proposed change does not involve a

physical alteration of the unit (i.e., no new

or different equipment will be installed), nor

will the function of equipment be changed.

The change will allow for a one time

performance of a natural circulation test in

Mode 3 which will provide useful data on

the natural circulation capabilities of the new

Babcock and Wilcox International (BWI)

steam generators that were recently installed

at Catawba Unit 1. The test data will be

utilized to validate analysis and simulator

models. Plant operators will also receive

valuable experience from performance of the

test. The test will be conducted using written

and approved procedures. An Emergency

procedure (EP/1/A/5000/ECA-0.1) is also

available to the Operators for this test. This

test is bounded by the Loss of NonEmergency AC Power to the Station

Auxiliaries event in Section 15.2.6 of the

Catawba UFSAR. For these reasons, the

planned natural circulation test will not

create the possibility of a new or different

type of accident from any previously

evaluated.

3) The activity does not involve a

significant reduction in the margin of safety.

Margin of safety is associated with

confidence in the ability of the fission

product barriers (the fuel and fuel cladding,

the Reactor Coolant System pressure

boundary, and the containment) to limit the

level of radiation doses to the public. As

demonstrated by the bounding UFSAR

analysis in Section 15.2.6, none of the fission

product barriers are adversely impacted by

the proposed one-time change. The proposed

change does not alter the manner in which

safety limits, limiting safety system setpoints,

or limiting conditions for operation are

determined. For these reasons, the activity

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the proposed

amendments involve no significant

hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Florida Power and Light Company,

Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of amendment request: February

24, 1997, as supplemented on April 24,

1997.

Description of amendment request:

The licensee proposed changes to

Technical Specification Section 6.9.1.7,

Core Operating Limits Report, to reflect

use of the Westinghouse Best Estimate

Large Break Loss-of-Coolant Accident

(LOCA) methodology for large break

LOCA analysis, including supporting

documents.

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:

Question 1 Does the proposed license

amendment involve a significant increase in

the probability or consequences of an

accident previously evaluated?

The plant conditions assumed in the

analysis are bounded by the design

conditions for all equipment in the plant.

Therefore, there will be no increase in the

probability of a Loss of Coolant Accident
30632 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

(LOCA). The consequences of a LOCA are not

being increased. That is, it is shown that the

emergency core cooling system is designed so

that its calculated cooling performance

conforms to the criteria contained in 10 CFR

50.46 paragraph (b). No other accident is

potentially affected by this change.

Therefore, neither the BiWeekly probability

nor the consequences of an accident

previously evaluated is increased due to the

proposed change.

Question 2 Does the proposed license

amendment create the possibility of a new or

different kind of accident from any accident

previously evaluated?

No new modes of plant operation are being

introduced. The parameters assumed in the

analysis are within the design limits of

existing plant equipment. All plant systems

will perform as designed in response to a

potential accident. Therefore, the proposed

license amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Question 3 Does the proposed amendment

involve a significant reduction in the margin

of safety?

The analysis in support of the proposed

license amendment realistically models the

expected response of the Turkey Point Units

3 & 4 nuclear core during a postulated LOCA.

Uncertainties have been accounted for as

required by 10 CFR 50.46. A sufficient

number of loss of coolant accidents with

different break sizes, different break locations

and other variations in properties have been

calculated to provide assurance that the most

severe postulated loss of coolant accidents

were analyzed. It has been shown by the

analysis that there is a high level of

probability the criteria contained in 10 CFR

50.46 paragraph (b) would not be exceeded.

Therefore, the proposed amendment does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408-

0420

NRC Project Director: Frederick J.

Hebdon

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: February

17, 1997 as revised May 1, 1997.

Description of amendment request:

The proposed amendment would

change the Crystal River Unit 3 (CR-3)

Technical Specifications (TS) to

implement 10 CFR Part 50, Appendix J,

‘‘Primary Reactor Containment Leakage

Testing for Water-Cooled Reactors,’’

Option B. This option allows to change

from prescriptive testing requirements

to performance-based testing

requirements based on the leakage rate

testing history of the containment and

components. The proposed TS changes

include revision to TS 3.6.1, 3.6.3, and

addition of ‘‘Containment Leakage Rate

Testing Program’’ to TS 5.0. The

licensee did not propose any deviations

from methods approved by the

Commission and endorsed in the

applicable regulatory guide. This notice

supersedes the previous notice dated

February 28, 1997 (62 FR 9214)

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

Does not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The TS amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to the TS are to

implement Option B of 10 CFR 50, Appendix

J, at CR-3. The proposed changes will result

in increased intervals between containment

leakage tests based on the leakage rate testing

history. The proposed changes do not involve

a change to the plant design or operation and

does not change the testing methodology.

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’ provides

the technical basis of 10 CFR 50, Appendix

J, Option B. NUREG-1493 contains a detailed

evaluation of the expected leakage from

containment and the associated

consequences. The increased risk due to

increasing the intervals between containment

leakage tests was also evaluated. The NUREG

used a statistical approach to determine that

the increase in the expected dose to the

public due to decreasing the testing

frequency is extremely low. NUREG-1493

also concluded that a small increase is

justifiable in comparison to the benefits from

decreasing the testing frequency. The

primary benefit is in the reduction in

occupational radiation exposure.

Criterion 2

Does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The TS amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS amendment incorporates

the performance-based testing approach

authorized by 10 CFR 50 Appendix, J, Option

B. Decreasing the testing frequency allowed

by this change does not involve a change to

plant design or operation. Safety related

equipment and safety functions are not

altered as a result of this change. Decreasing

the testing frequency does not affect testing

methodology. As a result, the proposed

change does not affect any of the parameters

or conditions that could contribute to the

initiation of any accidents.

Criterion 3

Does not involve a significant reduction in

the margin of safety.

This TS amendment does not involve a

significant reduction in the margin of safety.

The proposed TS amendment does not

change the methodology of the containment

leakage rate testing program or program

acceptance criteria. The proposed TS change

does affect the frequency of containment

leakage rate testing. With an increased

interval between tests, a small possibility

exists that an increase in leakage could go

undetected for a longer period of time. Based

on the operational experience at CR-3, it has

been demonstrated that the leak-tightness of

the containment building has consistently

been significantly below the allowable

leakage limit. Adequate controls are in place

to ensure that required maintenance and

modifications are performed.

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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Attorney for licensee: R. Alexander

Glenn, Corporate Counsel, Florida

Power Corporation, MAC - A5A, P. O.

Box 14042, St. Petersburg, Florida

33733-4042

NRC Project Director: Frederick J.

Hebdon

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: March

27, 1997, as supplemented April 3, and

May 1, 1997.

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) for the

Crystal River Nuclear Plant Unit 3 (CR3)

relating to the Once Through Steam

Generator’s (OTSG’s) tube inspection

acceptance criteria. Specifically, the

licensee proposed to:

(1) revise TS 3.4.12 (d) to specify 150

gallons per day limit on primary-tosecondary leakage through either OTSG;

(2) add TS 5.6.2.10.2 e. to define

inspection requirements and disposition

criteria for applicable tubes in the ‘‘B’’

OTSG first span;
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30633

(3) revise TS 5.6.2.10.4.a.7 to define

‘‘pit-like Intergranular attack

indications≥

(4) revise TS 5.6.2.10 and 5.7.2 to

delete requirements that were specific to

the interim tube plugging criteria

applicable until Refuel 11.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

FPC Response:

No. The CR-3 components addressed by

this proposed change are the Once Through

Steam Generators (OTSGs), identified by

plant tagging procedures as RCSG-1A and

RCSG-1B. The OTSGs are straight tube,

straight shell heat exchangers which allow

for heat removal and the subsequent

production of steam as a result of heat

transfer from the primary side reactor coolant

to the secondary side feedwater. Proposed

changes are; retaining reduced primary-tosecondary leak rates approved previously for

one cycle only, returning inspection result

reporting requirements to those previously

implemented, and establishing new

inspection requirements for the ‘‘B’’ OTSG.

Sunset clauses are being removed from pages

containing requirements effective for one

refueling outage and subsequent operating

cycle only.

Based on review of Chapter 14 of the CR3 Final Safety Analysis Report (FSAR), FPC

performed analyses to assess the

consequences of a steam generator tube

rupture event, including the complete

severing of a steam generator tube. The

analyses concluded that CR-3 was

sufficiently designed to ensure that, in the

event of a steam generator tube rupture, the

radiological doses would not exceed the

allowable limits prescribed by 10 CFR 100,

and would not result in additional tube

failures and further degradation of the reactor

coolant pressure boundary.

Retaining the present primary-to-secondary

leakage limit (LCO 3.4.12, RCS Operational

Leakage) that was previously approved for

the current operating cycle will continue to

provide assurance that should a significant

leak occur, it would be detected and the

plant will be shut down in a timely manner

to reduce the likelihood of a potential tube

rupture. This value of primary-to-secondary

leakage applicable to both OTSGs is

conservative relative to existing safety

analyses and would result in lower doses

than currently calculated and found

acceptable. Removing reporting requirements

specific to use of alternate flaw sizing criteria

approved for Refueling Outage 10 only, and

returning to previous reporting requirements

applicable to both OTSGs, has no effect on

operating plant safety. These requirements

are administrative only and do not affect

steam generator inspection or disposition of

inspection results.

The proposed change to the ‘‘B’’ OTSG

inspection criteria establishes that future

inspections will include 100% inspection of

the first span of specific tubes which are

known to have indications of degradation.

The degradation of these tubes is attributed

to a common non-random mechanism.

The results of inspections of these tubes

will be dispositioned using the same criteria

as all other OTSG tubes for determination of

the need for plugging or sleeving. Therefore,

the proposed change will not increase the

probability or consequence of an accident

previously evaluated as all tubes degraded

beyond acceptable limits will be subject to

consistent corrective actions.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

FPC Response:

No. The purpose of OTSG tube inspection

is to identify tubes that may have a higher

potential for failure due to degradation that

results in a reduced ability to withstand

operating conditions. Neither the type of

inspection of OTSG tubes nor the process for

performing inspections will be changed by

this amendment. Consistent criteria will be

applied to disposition inspection results and

consistent corrective actions will be taken for

tubes that exceed this criteria. Retaining the

lower leakage limit is conservative relative to

existing analyses. Changes to revise

requirements for reporting inspection results,

and remove ‘‘sunset’’ clauses addressing the

applicability of License Amendment 154

until Refueling Outage 11 only, do not alter

the design or operation of the OTSGs.

Therefore, no new or different kind of

accident will be created as a result of these

changes.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in margin of

safety?

FPC Response:

No. The analyses that have been performed

on the effects of OTSG tube failures, as

reported in the CR-3 FSAR, have

demonstrated that internal and offsite

consequences are within allowable limits.

This change will not alter the acceptance

criteria for inspection results. Since this

change will assure that a group of tubes with

existing first span pit-like inter-granular

attack indications are inspected each

inspection period, the likelihood of detecting

active degradation, as well as the probability

of repairing degraded tubes prior to the

degradation resulting in a through-wall

opening or tube rupture, is increased.

Retaining the currently accepted primary-tosecondary leakage limit continues to provide

assurance that should a significant leak

occur, it would be detected and the plant will

be shut down in a timely manner to reduce

the likelihood of a potential tube rupture,

thereby maintaining or improving the

existing margin of safety. Changes to revise

requirements for reporting inspection results,

and remove ‘‘sunset’’ clauses addressing the

applicability of License Amendment 154

until Refueling Outage 11 only, do not alter

the design or operation of the OTSGs.

Therefore, these changes will not involve a

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Attorney for licensee: R. Alexander

Glenn, Corporate Counsel, Florida

Power Corporation, MAC-A5A, P.O. Box

14042, St. Petersburg, Florida 33733-

4042

NRC Project Director: Frederick J.

Hebdon

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: May 8,

1997

Description of amendment request:

The proposed amendment incorporates

additional analytical methods, GPU

Nuclear Topical Reports, TR-078, TR087, TR-091, and TR-092P, previously

approved by the NRC, to Technical

Specifications (TS) Section 6.9.5.2.

These Topical Reports will be utilized

by GPU Nuclear to perform core reload

design analysis for the Three Mile

Island, Unit 1 (TMI-1) Facility. TS

6.9.5.2 is also being editorially revised

to relocate the existing note that the

current revision level shall be specified

in the Core Operating Limits Report

(COLR) such that it applies to the

additional Topical Reports, as well as

BAW-10179 P-A.

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 (SHC), which is presented

below:

GPU Nuclear has determined that this

Technical Specification Change Request

poses no significant hazards as defined by 10

CFR 50.92.

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The proposed change to reference

the analytical methodologies specified in

GPU Nuclear Topical Reports TR-078, TR087, TR-091,and TR-092 use[d] in TMI-1 core

reload design analysis is considered

administrative since these Topical Reports
30634 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

have been reviewed and approved by the

NRC for use at TMI-1.

Therefore, the proposed change does not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed change to reference

NRC-approved GPU Nuclear Topical Reports

TR-078, TR-087, TR-091, and TR-092P will

continue to ensure that approved methods

and criteria are used to establish core

operating limits.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety. The proposed change to reference

NRC-approved GPU Nuclear Topical Reports

TR-078, TR-087, TR-091, and TR-092P

maintains existing margins of safety since

approved methods and criteria are still used

to establish core operating limits.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Patrick D.

Milano, Acting

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of amendment request: May 8,

1997

Description of amendment request:

The proposed amendment would

modify the minimum accuracy stated in

Technical Specification (TS) Table 3.3-

8, ≥Meteorological Monitoring

Instrumentation,’’ for the instruments

used to measure wind speed and air

temperature - delta T. TS Bases Section

3/4.3.3.4 would also be modified to

reflect the proposed changes to TS Table

3.3-8.

Regulatory Guide 1.23 (Safety Guide

23), ‘‘Onsite Meteorological Programs,’’

dated March 17, 1972, provides

recommended instrument accuracies for

meteorological instrumentation. The

proposed minimum instrument

accuracies for the air temperature - delta

T and the wind speed (only when the

wind speed is greater than 5 miles per

hour) do not meet the recommended

accuracies of Regulatory Guide 1.23.

However, margin is included to account

for uncertainties.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed changes modify the accuracy

requirements for the instruments which are

used to measure wind speed and air

temperature - delta T. The data obtained from

the meteorological instrumentation would be

used to: a) estimate the public dose following

routine or accidental releases of airborne

radioactivity, b) make decisions regarding

actions to protect the public in the event of

an accident involving a release of airborne

radioactivity, and c) establish radiological

dispersion parameters to determine

radiological doses in design basis accident

calculations.

The proposed minimum instrument

accuracy requirements are more than

sufficient to meet the purposes denoted

above. The meteorological parameters

measurement uncertainties insignificantly

affect the results when compared to the

accuracies of the source term estimates,

meteorological dispersion models, dose

models, and meteorological forecasting.

Therefore, there is no impact on the

consequences (offsite doses) associated with

previously evaluated accidents.

The proposed changes do not alter the way

any structure, system, or component

functions, do not alter the manner in which

the plant is operated, and do not have any

impact on the protective boundaries and

safety limits for the protective boundaries.

Therefore, the proposed changes do not

impact the probability of any previously

evaluated accidents.

Thus, the license amendment request does

not impact the probability of an accident

previously evaluated nor does it involve a

significant increase in the consequence of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes modify the accuracy

requirements for the instruments which are

used to measure wind speed and air

temperature - delta T. The data provided by

these instruments assist in responding to a

design basis accident which may involve a

release of airborne radioactivity. The

instruments are used for post accident

monitoring and serve a passive role; they

cannot initiate or mitigate any accident.

The proposed changes do not alter the way

any structure, system, or component

functions and do not alter the manner in

which the plant is operated. They do not

introduce any new failure modes.

Thus, the license amendment request does

not create the possibility of a new or different

kind of accident from any previously

analyzed.

3. Involve a significant reduction in a

margin of safety.

As discussed above, the proposed changes

modify the accuracy requirements for the

instruments which are used to measure wind

speed and air temperature - delta T which

could impact the radiological dispersion

coefficient used to determine radiological

doses in design basis accident calculations.

However, the differences in the instrument

accuracies and the Regulatory Guide 1.23

requirements have been determined not to

significantly affect the dispersion

coefficients. Thus, there is no significant

impact on offsite doses associated with

previously analyzed accidents. Therefore,

there is no significant reduction in the

margin of safety for the design basis accident

analysis.

Thus, the license amendment request does

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Deputy Director: Phillip F. McKee

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: April 14,

1997

Description of amendment request:

Technical Specification 3.4.9.3.a

requires two relief valves be operable to

protect the reactor coolant system from

overpressurization when any reactor

coolant system cold leg is less than

350—F. The proposed amendment

revises the setpoint of the residual heat

removal suction relief valves.

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
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30635

consideration, which is presented

below:

NNECO has reviewed the proposed change

in accordance with 10CFR 50.92 and has

concluded that the change does not involve

a significant hazards consideration (SHC).

The bases for this conclusion is that the three

criteria of 10CFR 50.92(c) are not satisfied.

The proposed change does not involve a SHC

because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to Technical

Specification 3.4.9.3 to decrease the setpoint

of the residual heat removal suction relief

valves from 450 psig [plus or minus] 3% to

440 psig [plus or minus] 3% ([greater than or

equal to] 426.8 psig and [less than or equal

to] 453.2 psig) is consistent with the design

capabilities and system requirements of the

relief valves and the relief valves are not

credited in previously evaluated accidents.

Therefore, the proposed change 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.

The proposed change to Technical

Specification 3.4.9.3 to decrease the setpoint

of the residual heat removal suction relief

valves from 450 psig [plus or minus] 3% to

440 psig [plus or minus] 3% ([greater than or

equal to] 426.8 psig and [less than or equal

to] 453.2 psig) does not change the operation

of the residual heat removal system, reactor

coolant system or any system component

during normal or accident evaluations. The

proposed change to the setpoint of the

residual heat removal suction relief valves

from 450 psig [plus or minus] 3% to 440 psig

[plus or minus] 3% ([greater than or equal to]

426.8 psig and [less than or equal to] 453.2

psig) also ensures protection of the reactor

coolant system against cold

overpressurization transients in accordance

with the requirements of 10CFR50, Appendix

G.

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 change to Technical

Specification 3.4.9.3 to decrease the setpoint

of the residual heat removal suction relief

valves from 450 psig [plus or minus] 3% to

440 psig [plus or minus] 3% ([greater than or

equal to] 426.8 psig and [less than or equal

to] 453.2 psig) provides an acceptable

allowance between the maximum relief valve

setpoint ([less than or equal to] 453.2 psig)

and 10CFR50, Appendix G requirements. The

proposed change to the setpoint provides

sufficient allowance between the minimum

relief valve setpoint ([greater than or equal to]

426.8 psig) and reactor coolant system

pressure when residual heat removal system

is unisolated from the reactor coolant system

to minimize the probability of an inadvertent

residual heat removal system relief valve

opening.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

change 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, CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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: April 28,

1997

Description of amendment request:

Technical Specification Surveillances

4.1.2.3.1, 4.1.2.4.1, 4.5.2.f, and 4.5.2.h

require the charging and safety injection

pumps to be tested on a periodic basis

and after modifications that alter

subsystem flow characteristics. The

proposed amendment would increase

the required differential pressure at

recirculation flow for the safety

injection and centrifugal charging

pumps; decrease the required individual

safety injection and centrifugal charging

pump injection line flow rate; increase

the allowed individual safety injection

pump total flow rate; and make editorial

changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

changes in accordance with 10CFR50.92 and

has concluded that the changes do not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

the three criteria of 10CFR50.92(c) are not

satisfied. The proposed changes do not

involve [an] SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed changes to Technical

Specification Surveillances 4.1.2.3.1,

4.1.2.4.1, and 4.5.2.f to increase the required

discharge pressure for the centrifugal

charging pumps on recirculation flow during

surveillance testing from [greater than or

equal to] 2411 psid to [greater than or equal

to] 5676 ft (2464 psid) are consistent with

centrifugal charging pump design

requirements. The change in the referenced

units from differential pressure measured in

psid to total head measured in feet for the

centrifugal charging pumps and safety

injection pumps during surveillance testing

is an administrative change.

The proposed changes to Technical

Specification Surveillance 4.5.2.f to increase

the required discharge pressure for the safety

injection pumps on recirculation flow during

surveillance testing from [greater than or

equal to] 1348 psid to [greater than or equal

to] 3240 ft (1406 psid) are consistent with

safety injection pump design requirements.

The proposed changes to Surveillance

4.5.2.h: to decrease the required individual

centrifugal charging pump injection line flow

rate sum from [greater than or equal to] 339

gpm to [greater than or equal to] 310.5 gpm,

decrease the required individual safety

injection pump injection line flow rate sum

from [greater than or equal to] 442.5 gpm to

[greater than or equal to] 423.4 gpm, increase

the required individual safety injection Pump

A total flow rate from [less than or equal to]

670 gpm to [less than or equal to] 675 gpm,

and increase the required individual safety

injection Pump B total flow rate from [less

than or equal to]

650 gpm to [less than or equal to] 675 gpm

are consistent with centrifugal charging

pump and safety injection pump design

requirements.

The proposed changes are consistent with

equipment design requirements and

performing surveillance testing does not

involve a significant increase in the

probability of an accident previously

evaluated.

The proposed changes to the surveillance

testing of the centrifugal charging pumps and

safety injection pumps provide the necessary

assurance that the pumps will function

consistent with the flows used in the

accident analyses and does not involve a

significant increase in the consequence of an

accident previously evaluated.

Therefore, the proposed changes do 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.

The proposed changes to the surveillance

testing of the centrifugal charging pumps and

safety injection pumps do not change the

operation of the centrifugal charging or safety

injection systems or any of its components

during normal or accident evaluations. The

increase in the allowed maximum safety

injection pump flow does not impact the cold

overpressure accident analysis.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.
30636 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

3. Involve a significant reduction in a

margin of safety.

The proposed changes to Technical

Specification Surveillances 4.1.2.3.1,

4.1.2.4.1 and 4.5.2.f to increase the required

discharge pressure for the centrifugal

charging pumps on recirculation flow during

surveillance testing from [greater than or

equal to] 2411 psid to [greater than or equal

to] 5676 ft (2464 psid) provides an acceptable

margin between the required surveillance

and design pump performance to provide

assurance that the pumps will operate

consistent with the assumptions of the

accident analysis.

The proposed changes to Technical

Specification Surveillance 4.5.2.f to increase

the required discharge pressure for the safety

injection pumps on recirculation flow during

surveillance testing from [greater than or

equal to] 1348 psid to [greater than or equal

to] 3240 ft (1406 psid) provides an acceptable

margin between the required surveillance

and design pump performance to provide

assurance that the safety injection pumps

will operate consistent with the assumptions

of the accident analysis.

The proposed changes to Surveillance

4.5.2.h to decrease the required individual

centrifugal charging pump injection line flow

rate sum from [greater than or equal to] 339

gpm to [greater than or equal to] 310.5 gpm,

decrease the required individual safety

injection pump injection line flow rate sum

from [greater than or equal to] 442.5 gpm to

[greater than or equal to] 423.4 gpm, increase

the required individual safety injection Pump

A total flow rate from [less than or equal to]

670 gpm to [less than or equal to] 675 gpm

and increase the required individual safety

injection Pump B total flow rate from [less

than or equal to] 650 gpm to [less than or

equal to] 675 gpm are consistent with the

assumptions of the accident analysis. The

maximum allowed safety injection flow is

consistent with the vendor recommendation

for maximum continuous runout flow. Also,

the safety injection

pumps are disabled during specific normal

operating modes, consistent with the

assumptions of the accident analysis, to

ensure that they can not be an injection

source when the cold overpressure system is

required to be operable and thus the increase

in maximum safety injection pump flow does

not affect the cold overpressure accident

analysis.

The change in the referenced units in

Technical Specification Surveillances

4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f from

differential pressure measured in psid to total

head measured in feet for the centrifugal

charging pumps and safety injection pumps

during surveillance testing is an

administrative change.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

changes do 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, CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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: April 28,

1997

Description of amendment request:

Technical Specification 3.7.6 requires

that flood protection be provided for the

service water pump cubicles and

components when the water level

exceeds a specific value. The proposed

amendment (1) adds the closing of the

service water pump cubicle sump drain

valves, (2) revises the wording of the

action statement to be consistent with

the limiting condition for operation, and

(3) revises the associated Bases section.

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 change

in accordance with 10CFR50.92 and has

concluded that the change does not involve

a significant hazards consideration (SHC).

The bases for this conclusion is that the three

criteria of 10CFR50.92(c) are not satisfied.

The proposed change does not involve [an]

SHC because the change would not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed changes to Technical

Specification 3.7.6 identify additional

manual actions to be performed to provide

external lood protection for the service water

pump cubicles in the event of high water

level (13 ft MSL) [mean sea level]. The

cubicle sump drain valves which are to be

closed are part of a modification which

installed a drain line from the sump of each

cubicle to the intake bay in order to provide

a passive means of removing internal leakage

from the cubicle. The cubicle sump drain

valves are normally maintained in the open

position.

The drain valves meet the intent of RG

[Regulatory Guide] 1.59 for ‘‘hardened

protection’’ and RG 1.102 for ‘‘incorporated

barriers’’ in a manner similar to that of the

cubicle watertight doors. RG 1.59 states that

hardened protection ‘‘must be passive and in

place, as it is to be used for flood protection,

during normal plant operation’’. RG 1.102

states that ‘‘the plant should be designed and

operated to keep doors necessary for flood

protection closed during normal operation’’.

The Response to FSAR [Final Safety Analysis

Report] Question No. 240.9 established the

acceptability of the practice of maintaining

one service water pump cubicle watertight

door open and the other door closed during

normal operations.

The proposed change in the action

statement to initiate action when water level

is exceeding 13 feet MSL rather than at 13

feet MSL is a clarification only which

provides consistency between the limiting

condition for operation and the action

statements.

Therefore, the proposed changes do 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.

The proposed changes to Technical

Specification 3.7.6 identify additional,

simple to perform manual actions to provide

external flood protection for the service water

pump cubicles.

The proposed change in the action

statement to initiate action when water level

is exceeding 13 feet MSL rather than at 13

feet MSL and the proposed changes to the

bases are considered clarifications.

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 to Technical

Specification 3.7.6 identify additional,

simple to perform manual actions to provide

external

flood protection for the service water pump

cubicles in the event of high water level (13

ft MSL). The plant modification which made

these additional actions necessary was made

to provide for improved internal flood

protection.

The proposed change in the action

statement to initiate action when water level

is exceeding 13 feet MSL rather than at 13

feet MSL and the proposed changes to the

bases are considered clarifications.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

change 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,
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30637

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, CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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: May 1,

1997

Description of amendment request:

Technical Specifications 3/4.8.2.2 and

3/4.8.3.2 specify which electrical power

systems are required to be operable in

Modes 5 and 6. The proposed

amendment would clarify the

requirements by identifying the specific

equipment required and their

alignments in Modes 5 and 6.

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

changes in accordance with 10CFR 50.92 and

has concluded that the change does not

involve a significant hazards consideration

(SHC). The bases for this conclusion is that

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

satisfied. The proposed changes do not

involve [an] SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed change to Technical

Specification 3/4.8.2.2 to replace the wording

‘‘As a minimum, one 125 volt battery bank

and its associated full capacity charger’’ to

‘‘As a minimum, one Train

(A or B) of batteries and their associated

full capacity chargers’’ will increase the

required battery banks operable from one to

two.[≥]

This change is being proposed to resolve

an inconsistency with Technical

Specification 3/4.8.3.2 which currently

requires two battery banks energized in

modes 5 and 6.

The proposed change to...Technical

Specifications 3/4.8.2.2 and 3/4.8.3.2 to

identify the specific equipment required and

its alignment during modes 5 and 6 is being

proposed to reduce the vagueness in the

present Technical Specifications. This

proposed change will specify the equipment

required operable for the electrical

distribution systems during modes 5 and 6.

These proposed changes are considered

administrative and do not alter the manner

in which any system or component is

operated or expected to respond during an

accident. Therefore, the proposed changes do

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.

The proposed changes to Technical

Specification 3/4.8.2.2 to increase the

required battery banks operable from one to

two and to reword Technical Specifications

3/4.8.2.2 and 3/4.8.3.2 to identify the specific

equipment required operable during modes 5

and 6 do not alter the manner in which any

system or component is operated or expected

to respond during normal or accident

conditions.

Therefore, the proposed changes do 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 to Technical

Specification 3/4.8.2.2 to increase the

required battery banks operable from one to

two is being proposed to resolve an

inconsistency with Technical Specification

3/4.8.3.2 which currently requires two

battery banks energized in modes 5 and 6.

This is considered an administrative change.

The proposed changes to...Technical

Specifications 3/4.8.2.2 and 3/4.8.3.2 are

being proposed to reduce the vagueness in

the present technical specifications by

identifying the specific equipment required

operable during modes 5 and 6. The change

will provide a greater level of assurance that

the electrical distribution systems will be

correctly aligned and surveilled. This is also

considered an administrative change.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

changes to 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, CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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: May 5,

1997

Description of amendment request:

Technical Specification Surveillance

4.8.4.1 requires periodic testing of lower

voltage circuit breakers for all

containment penetration conductor

overcurrent protective devices. The

proposed amendment would modify the

requirements for determining the

operability of lower voltage circuit

breakers by using the manufacturer’s

curve of current versus time to test

delay trip elements, clarify the use of

two pole in series testing, and expand

the Bases description of the testing.

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 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The bases for this conclusion is that

the three criteria of 10CFR50.92(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.

The proposed change to Technical

Specification Surveillance 4.8.4.1 to modify

the requirements for determining the

operability of lower voltage circuit breakers

by using the manufacture’s curve of current

versus time to test long time and short-time

delay trip elements will not change the

requirement that periodic testing be

performed to determine breaker operability.

The circuit breaker testing is consistent with

the design of the components and performing

surveillance testing does not involve a

significant increase in the probability of an

accident previously evaluated. The proposed

change will provide assurance that the

breakers will perform consistent with

accident analyses and does not involve a

significant increase in the consequence of an

accident previously evaluated.

The proposed change to the surveillance to

modify the wording associated with the use

of two pole in series testing to determine

Molded Case Circuit Breaker (MCCB)

operability following the failure of [an]

MCCB to pass a single pole test was

previously approved in License Amendment

No. 13. The modified wording clarifies the

testing by specifically stating in the

surveillance that the two pole in series test

determines MCCB operability. This is

considered an administrative change.

The proposed change to expand the

description of the long-time and short-time

delay trip elements testing in the Bases

Section is also considered an administrative

change.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.
30638 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change to use a curve of

current versus time instead of the description

in Technical Specification Surveillance

4.8.4.1 of the [] long-time and short-time

delay trip element testing does not alter the

design, operation, or maintenance of the

lower voltage circuit breakers.

The proposed change to the surveillance to

modify the wording associated with the use

of two pole in series testing to determine

MCCB operability and the expanded

description of the long-time and short-time

delay elements testing in the Bases Section

are considered administrative changes.

Therefore, the proposed changes do 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 current wording of Technical

Specification Surveillance 4.8.4.1 requires

testing of long-time delay trip elements with

a current value of exactly 300% of the pickup

setting and short-time delay trip elements

with a current value of exactly 150% of the

pickup setting. The testing [cannot] be

performed at exact values. Circuit breaker

manufactures develop a curve of current

versus time for each breaker type that

specifies the allowable time to trip for a

specified current. Using the curve for a given

breaker type, the operability of a circuit

breaker can be verified by inserting a given

current and verifying that the breaker trips

within the allowable time delay band width

for that current. Testing by the industry is

typically performed at approximately 300%

of the pickup setting for long-time delay trip

elements and approximately 150% of the

pickup setting for short-time delay trip

elements. The proposed change to the

surveillance to modify the requirements for

determining the operability of circuit

breakers by using the manufacturer’s curve of

current versus time to test delay trip

elements will continue to provide assurance

that lower voltage circuit breakers for all

containment penetration conductor

overcurrent protective devices will operate

consistent with the assumptions of the

accident analysis.

The proposed change to the surveillance to

modify the wording associated with the use

of two pole in series testing to determine

MCCB operability and the expanded

description of the long-time and short-time

delay trip elements testing in the Bases

Section are considered administrative

changes.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

changes do 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, CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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: May 5,

1997

Description of amendment request:

Technical Specification Surveillance

4.5.2.b.1 requires that the emergency

core cooling system (ECCS) piping be

verified full of water at least once per 31

days. The proposed amendment would

revise the surveillance to exempt the

operating charging pump(s) and

associated piping from the requirement

to be verified full of water and move the

description of the verification method

from the surveillance to the Bases

section.

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 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The bases for this conclusion is that

the three criteria of 10CFR50.92(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.

The proposed change to Technical

Specification Surveillance 4.5.2.b.1 to

exempt the operating centrifugal charging

pump(s) and associated piping from the

requirement to be vented will not effect the

requirement the ECCS piping be full of water.

An operating centrifugal charging pump and

the associated piping is self venting and

cannot develop voids and pockets of

entrained gases. This change is consistent

with the design of the charging system and

ensuring that ECCS piping is full of water

does not involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed change Technical

Specification Surveillance 4.5.2.b.1 to move

and expand the description of the venting

method from the surveillance to the Bases

Section are considered administrative

changes.

Therefore, the proposed changes do 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.

The proposed change to exempt the

operating centrifugal charging pump(s) and

associated piping from the requirement to be

periodically vented by crediting its self

venting capabilities does not change the

operation of the charging system or any of its

components during normal or accident

evaluations.

The proposed changes to move and expand

the description of the venting method from

the surveillance to the Bases Section are

considered administrative changes.

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 change to Technical

Specification Surveillance 4.5.2.b.1 to

exempt the operating centrifugal charging

pump(s) and associated piping from the

requirement to be manually vented by

crediting its self venting capabilities, is

consistent with the design of the charging

system. This proposed change continues to

ensure that ECCS piping is full of water and

thus, does not involve a significant reduction

in a margin of safety.

The proposed change to Technical

Specification Surveillance 4.5.2.b.1 to move

the description of the venting method from

thesurveillance to the Bases Section is

considered an administrative change.

Currently the surveillance identifies that

ECCS piping is to be verified full of water by

venting ECCS pump casings and accessible

discharge piping high points except for the

RSS [recirculation spray system] pump, RSS

heat exchanger and associated RSS piping

that are not maintained filled with water

during plant operation. The venting

description will be expanded when moved to

the bases to include an exclusion for the

above described operating centrifugal

charging pump(s) and associated piping and

the venting method used for nonoperating

centrifugal charging pumps. The centrifugal

charging pumps have top mounted suction

and discharge nozzles and do not have casing

vents. The pump manufacturer has indicated

that venting the pump suction pipe will

assure that the pump is full of water. This

venting of the nonoperating centrifugal

charging pumps is accomplished by using a

pump suction line test connection.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

change 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
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30639

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, CT 06141-0270

NRC Deputy Director: Phillip F. McKee

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: April 17,

1997

Description of amendment request:

This license amendment request revises

Technical Specification (TS) 2.12,

‘‘Control Room System,’’ to delete the

Limiting Conditions of Operation (LCO)

and associated surveillance for the

control room temperature and replace it

with an LCO and surveillance on the

control room air conditioning (A/C)

system. The remainder of TS 2.12 is

being rewritten consistent with the

requirements of the Combustion

Engineering Standard TS (NUREG-1432,

Rev. 1). In reviewing requirements for

refueling and shutdown operations,

additional TS improvement were

identified. Therefore, the definition

section, TS 2.1 ‘‘Reactor Coolant

System,’’ 2.6 ‘‘Containment System,’’

2.8 ‘‘Refueling Operations,’’ and

associated surveillance requirements are

proposed for revision to incorporate the

design basis requirements for refueling

operations and to correspond to

NUREG-1432.

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 will incorporate new

requirements for the control room air

conditioning system, control room filtration

system, and refueling operations. In addition,

the proposed change will ensure that the

Limiting Condition for Operations and

surveillance requirements are consistent with

the design basis of a fuel handling accident

as documented in the FCS Updated Safety

Analysis Report (USAR).

CONTROL ROOM SYSTEMS

The control room air conditioning (A/C)

system consists of two redundant A/C units,

VA-46A and VA-46B. Each unit has sufficient

capacity to meet the cooling requirements for

personnel and equipment inside the control

room envelope. Each A/C unit is equipped

with an air-cooled condenser located inside

a protective enclosure outdoors on the roof

of the Auxiliary Building. Each A/C unit’s

refrigerant compressor, air cooling coils, fans,

and dampers are located inside of the control

room envelope. Each unit has a waterside

economizer coil that allows air cooling with

Component Cooling Water (CCW). When

cooling water temperature is sufficiently low,

a temperature-activated valve at each A/C

unit allows cooling water flow through the

waterside economizer. This valve also diverts

flow away from the waterside economizer if

cooling water temperature is too high. The

air-operated CCW isolation valves to the A/

C units fail closed and are automatically

closed on a Ventilation Isolation Actuation

Signal (VIAS) to prevent CCW flow through

the waterside economizers in a post-accident

situation.

Technical Specification (TS) 2.12(1)

requires that the temperature within the

control room and control cabinets be

maintained below 120°F does not meet any

of the four criteria contained in 10 CFR 50.36

for inclusion in TS. However, the equipment

required to maintain this temperature, the

control room air conditioning system, meets

Criterion 3 of 10 CFR 50.36 in that the system

functions to mitigate a design basis accident

by maintaining the control room in a

habitable environment.

Therefore, it is proposed that this TS be

revised to delete the control room

temperature as a LCO and require that two

control room air conditioning trains be

operable when the reactor coolant

temperature is above 210°F. The design

temperature limits of instrumentation and

controls inside of the control room will be

maintained in the Basis Section of TS 2.12.

The allowed outage time for one train of

control room air conditioning is proposed as

30 days. This is consistent with Combustion

Engineering Standard TS 3.7.12 (NUREG1432 Rev. 1). In addition, the FCS

Probabilistic Risk Assessment model was

reviewed and validated a 30 day outage time

as being non-risk significant. The impact on

Core Damage Frequency (CDF) from a 30 day

LCO was based on the assumption that one

cooling unit was always inoperable and thus

under the LCO for an entire year. This allows

the analysis to consider unlimited entries

into the LCO and a full LCO duration for

each entry. Using this assumption, the

baseline (annually) CDF of 1.53E-5 would

increase by 21.6% to a frequency of 1.86E5. In accordance with EPRI’s ‘‘PSA

Applications Guide,’’ this small increase in

CDF can be classified as ‘‘non-risk

significant.’’

Specification 2.12(2)

Specification 2.12(2) requires that a

thermometer be in the control room at all

times. This instrumentation does not meet

any of the four criteria contained in 10 CFR

50.36 for inclusion in the FCS TS. Therefore,

the requirement is proposed for relocation to

the FCS USAR.

Specification 2.12(3)

Specification 2.12(3) requires that all areas

of the plant containing safety related

instrumentation be observed during hot

functional testing to determine local

temperatures and monitored during

operation if normal plant ventilation is not

available. It is proposed to delete this TS.

The requirement to monitor and determine

local temperatures during hot functional

testing was met during the initial startup

phase of FCS and is no longer applicable.

The requirement to monitor temperatures

within the plant during normal operation

does not meet any of the four criteria

contained in 10 CFR 50.36 for inclusion in

TS and therefore is being deleted.

The requirement to control temperatures

for safety related instrumentation and

controls, and initiate supplementary cooling

if required, is currently described in USAR

Section 9.10. These USAR requirements are

controlled by plant procedures. Any changes

to these requirements would require an

evaluation be conducted in accordance with

10 CFR 50.59.

Specification 2.12(4)

Specification 2.12(4) allows one control

room air filtration system to be inoperable for

7 days or a plant shutdown be commenced.

This specification does not state which

modes of operation it applies to.

Therefore, it is proposed to revise this

specification to require two trains of control

room air filtration systems to be operable

when the reactor coolant temperature is

above 210°F. The allowed outage time will be

maintained at 7 days and a total of 42 hours

will be allowed to take the plant to cold

shutdown. The 42 hour time period is

consistent with TS 2.0.1 which addresses

equipment outages in excess of what is

specifically allowed by individual

specifications.

The proposed changes for the control room

systems consist of providing additional

restrictions on operation of the control room

air filtration systems and control room air

conditioning system. These changes ensure

that equipment required to mitigate the

consequences of an accident are operable.

Therefore, the proposed changes do not

increase the probability or consequences of

an accident previously evaluated.

REFUELING OPERATIONS

The design bases of the fuel handling

accident and refueling operations were

reviewed and several inadequacies were

identified related to refueling operations.

Therefore, revisions are proposed for the TS

Definition section, TS 2.6 on containment

integrity, and TS 2.8 on refueling operations

to reflect NUREG-1432.

Definitions

Cold Shutdown Condition & Refueling

Shutdown Condition

The changes proposed for the definitions of

Cold Shutdown Condition, and Refueling

Shutdown Condition clarify these

definitions. The plant is in Cold Shutdown

when Tcold is less than 210°F, and the reactor

coolant is at least Shutdown Boron

Concentration but less than Refueling Boron

Concentration. Similarly, the definition for

Refueling Shutdown is clarified to apply

when Tcold is less than 210°F and the reactor
30640 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

coolant is at least Refueling Boron

Concentration. This change does not propose

any new operating modes but merely clarifies

when the definitions are applicable.

Core Alterations

The definition for Core Alterations is being

revised to reflect the requirements of

NUREG-1432. This revision deletes ‘‘any

component’’ from the definition and clarifies

that the components considered by this

definition are those that could affect

reactivity. In addition, the revision adds

nuclear fuel to the definition such that

movement of fuel within the reactor vessel

will be defined as a core alteration and not

a refueling operation.

Refueling Operations

The definition of Refueling Operations is

being revised to delete control element

assemblies (CEA) or startup sources from the

definition since these are items that are

included in the definition of Core

Alterations. Additionally, it is being revised

to specify that the definition is limited to

movement of irradiated fuel outside of the

reactor pressure vessel since fuel movement

inside the reactor vessel is included in the

definition of Core Alteration. Finally, a

clarification is being added to state that

suspension of refueling operations shall not

preclude completion of movement of

irradiated fuel to a safe, conservative

position.

In Operation

The definition of In Operation is being

revised to include the definition of operable.

This is a more conservative interpretation

than currently exists.

Specification 2.1 ‘‘Reactor Coolant

System’’

It is proposed to revise TS 2.1.1(3) to

include shutdown cooling requirements

when the reactor coolant system (RCS)

temperature is below 210°F with fuel in the

reactor and the reactor vessel head fully

tensioned. The definitions of Cold Shutdown

(Mode 4) and Refueling Shutdown (Mode 5)

contained in the TS make no distinction as

to the status of the reactor vessel head or RCS

temperature. The only difference between the

two defined modes is boron concentration.

Higher or lower boron concentration affects

shutdown margin but does not affect decay

heat load, which is the basis for this

specification.

Technical Specification 2.1.1(4) was

intended to address shutdown cooling

requirements during refueling operations.

However, this is already addressed in TS 2.8.

Therefore, it is proposed to delete TS 2.1.1(4)

and the exception since new specifications

addressing shutdown cooling loop

requirements during Mode 5 with fuel in the

reactor and with one or more reactor vessel

head closure bolts less than fully tensioned

are proposed for inclusion in TS 2.8

(Refueling Operations).

The associated statements supporting these

items in the Basis section are also proposed

for deletion. Prior to any reactor vessel head

closure bolts being loosened, TS 2.1.1 will be

applicable which will require two shutdown

cooling loops. As soon as a closure bolt is

loosened, the new proposed TS 2.8 would be

applicable which also requires two shutdown

cooling loops whenever there is less than 23

feet of water above the core. The

requirements of TS 2.1.1(3) are similar to

NUREG-1432, Specifications 3.4.7 and 3.4.8.

Specification 2.6 ‘‘Containment System’’

Currently, TS 2.6(1)c states that

containment integrity shall not be violated

when the reactor vessel head is removed if

the boron concentration is less than refueling

concentration. However, Specification 2.6(1)c

has no required actions and therefore, TS

2.0.1 must be entered when the LCO is not

met. In this situation, (reactor vessel head

removed), TS 2.0.1 is ineffective because the

plant would already be in Refueling

Shutdown. Thus, TS 2.6(1)c is proposed for

deletion.

Currently, Specification 2.6(1)d requires

that except for testing one control element

drive mechanism at a time, positive reactivity

changes shall not be made by CEA motion or

boron dilution unless containment integrity

is intact. Specification 2.6(1)d is proposed for

deletion as it is unnecessarily restrictive.

Specification 2.8.1(1) as proposed

eliminates the need for containment integrity

when the reactor is in Refueling Shutdown.

Specification 2.8.1(1) requires sufficient

shutdown margin to preclude a criticality

event and also prescribes actions to restore

the shutdown margin if necessary. Small

positive reactivity increases whether by CEA

motion or boron dilution will not cause a

criticality event due to the need to maintain

at least a 5% shutdown margin. Therefore,

the requirement to maintain containment

integrity is unnecessarily restrictive since a

criticality event cannot occur when a

shutdown margin of at least 5% exists.

Specification 2.8.1(1) is consistent with the

requirements of NUREG-1432, Specification

3.9.1.

A new specification (TS 2.8.2(1)) is

proposed that provides requirements for

containment closure during core alterations

and refueling operations inside of

containment. The design basis of the Fort

Calhoun Station does not require full

containment integrity during a fuel handling

accident. As stated in USAR Section 14.18,

the fuel handling accident does not take

credit for containment isolation. Therefore,

requiring full containment integrity is

inappropriate and requirements for

containment closure are proposed for

addition to TS 2.8 consistent with NUREG1432 Specification 3.9.2.

Specification 2.10.2 governs operation of

CEAs and monitoring of selected core

parameters. Specification 2.10.2 ensures (1)

adequate shutdown margin following a

reactor trip, (2) that the moderator

temperature coefficient (MTC) is within the

limits of the safety analysis, and (3) CEA

operation is within the limits of the setpoint

and safety analysis. Specification 2.10.2

ensures that the reactor will be maintained

sufficiently subcritical to preclude

inadvertent criticality and provides actions

(i.e., boration) to be taken to ensure that the

required shutdown margin is available. Thus,

TS 2.10.2 precludes the need for containment

integrity when the plant is in cold shutdown.

Specification 2.8 ‘‘Refueling Operations’’

It is proposed that TS 2.8 be rewritten to

reflect NUREG-1432. Currently, this

specification applies to any refueling

operation. However, no distinction is made

between refueling operations within

containment and refueling operations within

the spent fuel pool. In addition, several

initial assumptions of a fuel handling

accident are not addressed by the current TS

2.8.

Specification 2.8(1)

The current TS 2.8(1) is inadequate. This

specification requires that the equipment

hatch and one door in the Personnel Air Lock

be properly closed, and all automatic

containment isolation valves be operable or

at least one valve closed. The specification

does not define what is meant by a properly

closed equipment hatch; that information is

currently contained in the Basis of TS 2.1.1.

In addition, inclusion of all automatic

containment isolation valves instead of those

providing direct access to the outside

atmosphere is incorrect.

The containment isolation system is

defined in USAR Section 5.9.5 as those

devices actuated by a Containment Isolation

Actuation Signal (CIAS) or a Steam Generator

Isolation Signal (SGIS). This includes many

valves that have no design basis function

during a fuel handling accident. A CIAS is

initiated by a Containment Pressure High

Signal or a Pressurizer Pressure Low Signal.

Neither of these signals are required to be

operable during refueling operations as these

signals would/could not respond to a fuel

handling accident.

The correct requirements are specified in

TS 2.8(2) which only requires that closure be

initiated by the Ventilation Isolation

Actuation Signal (VIAS) for the containment

pressure relief, air sample, and purge system

valves. Due to these inadequacies, it is

proposed to delete TS 2.8(1) and replace it

with a new Specification 2.8.2(1) which is

consistent with NUREG-1432 Specification

3.9.3.

Specification 2.8(2)

It is proposed that TS 2.8(2) be deleted and

replaced by new Specifications 2.8.2(3) and

2.8.3(5). The requirement to maintain an

operable Ventilation Isolation Actuation

Signal with input from the containment

atmosphere gaseous and auxiliary building

exhaust stack gaseous radiation monitors is

consistent with current requirements and

required actions are consistent with NUREG1432, Specification 3.3.8. Radiation Monitor

RM-052 functions as a ‘‘swing’’ monitor, i.e.,

it can be aligned to monitor either

containment or the auxiliary building

exhaust ventilation stack. Radiation Monitor

RM-052 is powered by either MCC-3B1/AI40C (like RM-051) or MCC-4C2/AI-40D (like

RM-062).

Technical Specification 2.7, Electrical

System is not required to be applied when

the RCS is below 300°F. Above 300°F, TS 2.7

requires both 4160-VAC buses to be operable.

Thus, above 300°F the required radiation

monitors must be powered from independent

480-VAC buses supplied by independent

4160-VAC buses. During refueling outages,

bus alignments other than those used during

power operation are used to permit electrical

system maintenance and modifications.

In the loss of offsite power event, the

radiation monitor sample pumps and control

room HVAC units stop and will not restart
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30641

until the emergency diesel generators (EDGs)

reenergize the system. The fuel handling

equipment also stops and does not restart

when the EDGs reenergize the system, thus

minimizing the potential of a fuel handling

accident. When the EDGs reenergize the

buses, VIAS will operate as designed.

Therefore, when the RCS is below 300°F, the

required monitors need only be powered

from independent 480-VAC buses supplied

by a single 4160-VAC bus.

There is no need to assume that a fuel

handling accident occurs immediately

followed by a loss of offsite power. However,

in the unlikely event that this should occur,

there would be no effect on the site boundary

dose since VIAS is not credited in USAR

Section 14.18 (Fuel Handling Accident). In

this situation, when the EDGs reenergize the

buses, the control room HVAC units will

restart in the filtered air makeup mode and

the stack radiation monitor sample pump

will restart. However, the containment

radiation monitor sample lines remain

isolated preventing the restart of the monitor

sample pump after receipt of a VIAS.

Specification 2.8(3)

It is proposed that TS 2.8(3) be deleted.

This requirement does not meet any of the

four criteria contained in 10 CFR 50.36 for

inclusion in the TS. The requirement that

radiation levels in containment and the spent

fuel pool shall be monitored during refueling

operations will be incorporated into the FCS

USAR.

Specification 2.8(6)

It is proposed that TS 2.8(6) be deleted.

This requirement does not meet any of the

four criteria contained in 10 CFR 50.36 for

inclusion in the TS. The requirements that

direct communication between personnel in

the control room and at the refueling

machine shall be available whenever core

alterations are taking place will be

incorporated into the FCS USAR.

Specification 2.8(7)

It is proposed that TS 2.8(7) be deleted and

replaced with a new Specification 2.8.3(4).

The requirement to place the spent fuel pool

ventilation system in operation prior to

refueling operations is consistent with the

current TS. It is being clarified that this

specification only applies to refueling

operations in the spent fuel pool, and not

when conducting refueling operations inside

of containment. Additionally, it is being

clarified that TS 2.0.1 is not applicable to this

activity, as reactor operation is independent

of fuel movements in the spent fuel pool.

Specification 2.8(9)

The current Specification 2.8(9) is

inadequate. This specification requires a

minimum of 23 feet of water above the top

of the core. This does not meet the initial

conditions assumed in the fuel handling

accident as documented in USAR Section

14.18. USAR Section 14.18 assumes 23 feet

of water above where the fuel could land if

dropped. In order to meet this initial

condition, a minimum of 23 feet of water

above the reactor vessel flange is required, as

this is the highest point where a fuel bundle

could land if dropped. Procedures reflect the

requirement to maintain 23 feet of water

above the reactor vessel flange during

refueling operations. The proposed revision

is consistent with NUREG-1432,

Specification 3.7.16.

Specification 2.8(11)

The current specification is inadequate.

The specification provides restrictions on

storage of fuel in the spent fuel pool;

however, there are no required actions to

address situations when the specification is

not met. It is proposed that TS 2.8(11) be

deleted and replaced with a new

Specification 2.8.3(1) that requires that a

misloaded fuel assembly be moved

immediately. Additionally, it is being

clarified that TS 2.0.1 is not applicable to this

activity, as reactor operation is independent

of fuel movements in the spent fuel pool.

Specification 2.8(12)

It is proposed that TS 2.8(12) be deleted

and replaced with a new Specification

2.8.3(3). The requirement to maintain 500

ppm boron concentration in the spent fuel

pool whenever unirradiated fuel is stored

there is consistent with the current TS and

the required actions are consistent with

NUREG-1432, Specification 3.7.17.

Restriction on Movement of Irradiated Fuel

from the Reactor Core

The restriction on irradiated fuel

movement unless the core has been

subcritical for at least 72 hours if the reactor

has been operated at power levels above 2%

is proposed for relocation to the Bases of TS

2.8.2(2). This requirement does not meet any

of the four criteria contained in 10 CFR 50.36

for inclusion in the TS. This is consistent

with NUREG-1432, B 3.9.6.

Reactor Coolant System Boron

Concentration

Currently, there is no specification for

boron concentration. Refueling boron

concentration is included in the definition of

Mode 5. However, there are no required

actions to be taken if the boron concentration

should be below refueling concentration.

Therefore, it is proposed that a new

Specification 2.8.1(1) be incorporated

consistent with NUREG-1432, Specification

3.9.1.

Spent Fuel Pool Water Level

Currently, there is no specification for

spent fuel pool water level. The water level

of the spent fuel pool is an initial condition

assumed in USAR Section 14.18. It is

proposed that a new Specification 2.8.3(2) be

incorporated into TS 2.8, which is consistent

with NUREG-1432, Specification 3.7.16.

The proposed changes for the RCS and

containment during shutdown, and

requirements for refueling operations, consist

of providing additional restrictions on

operation, and changes to make the

requirements of the TS Limiting Conditions

for Operation consistent with the initial

conditions and assumptions of the fuel

handling accident as documented in USAR

Section 14.18. Therefore, the proposed

changes do not increase the probability or

consequences of an accident previously

evaluated.

SURVEILLANCE REQUIREMENTS

CONTROL ROOM

Specification 3.1, Table 3-3, Item 13.

Specification 3.1, Table 3-3, Item 13

requires that the thermometer in the control

room be compared with a calibrated

thermometer and replaced if out of tolerance

on a refueling

frequency. It is proposed that this

surveillance be deleted to be consistent with

deletion of the LCO requirement to maintain

a thermometer in the control room.

A new surveillance is proposed to verify

that the control room air conditioning system

has the capability to remove the assumed

heat load. This surveillance will ensure the

operability requirements for TS 2.12 are met.

The test and frequency is consistent with

NUREG-1432.

The air-operated CCW isolation valves to

the A/C units fail closed and are

automatically closed on a VIAS to prevent

CCW flow through the waterside

economizers in a post-accident situation.

These valves are currently tested in

accordance with TS 3.3 (FCS Inservice

Testing Program). Prior to the modification,

the valves were tested as fail-open valves. No

TS changes are necessary.

The control room air filtration system is

currently tested on a refueling frequency in

accordance with TS 3.2, Table 3-5, Item 10a.

No TS changes are necessary.

REFUELING OPERATIONS

Reactor Coolant Boron Concentration

During Refueling Operations

The Reactor Coolant System boron

concentration is currently sampled in

accordance with TS 3.2, Table 3-4, Item 1(e).

It is proposed to revise the frequency from

once per shift during refueling operations to

once per 3 days which is consistent with

NUREG-1432. As stated in the basis of TS 2.8

and USAR Section 14.18, the reactor cavity

is filled with over 200,000 gallons of borated

water prior to the start of refueling

operations. The requirements for sampling

the reactor coolant during the remainder of

Mode 5 is performed once per 3 days in

accordance with Table 3-4, Item 1(d). This

proposed change will make the sampling

consistent with the requirements of Item 1(d)

and NUREG-1432.

Spent Fuel Pool Boron Concentration

The spent fuel pool boron concentration is

currently sampled in accordance with TS 3.2,

Table 3-4, Item 5. It is proposed to revise the

frequency of the sampling to prior to

movement of unirradiated fuel in the spent

fuel pool and once per week whenever

unirradiated fuel is stored there to be

consistent with the requirements of the LCO.

Source Range Neutron Monitors

Currently, a channel check and calibration

of the wide range neutron monitors is

performed in accordance with TS 3.1, Table

3-1, Item 2.

Containment Penetrations

Currently, there is no surveillance to

determine the status of containment

penetrations during refueling operations.

Therefore, a new surveillance is proposed for

TS 3.2, Table 3-5 to verify the status of

required containment penetrations once per

7 days consistent with NUREG-1432.

The requirement of NUREG-1432 to verify

that the containment purge and exhaust

valves actuate to the isolation position on a

refueling frequency is currently tested as part

of the Containment Radiation High Signal

test required by TS 3.1, Table 3-2. Item 4.

Shutdown Cooling Loops

Currently, there is no surveillance

requirement to verify that the required
30642 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

shutdown cooling loops are operable and in

operation or to verify correct breaker lineup

for the shutdown cooling loop that is not in

operation. Therefore a new surveillance is

proposed to be incorporated into TS 3.2,

Table 3-5 consistent with NUREG-1432.

Refueling Water Level

Currently, there is no surveillance

requirement to verify the refueling water

level during refueling operations. Therefore,

a new surveillance is proposed for

incorporation into TS 3.2, Table 3-5

consistent with NUREG-1432.

Spent Fuel Pool Water Level

Currently, there is no surveillance

requirement to verify the spent fuel pool

water level during refueling operations.

Therefore, a new surveillance is proposed for

incorporation into TS 3.2, Table 3-5

consistent with NUREG-1432.

Spent Fuel Initial Enrichment/Burnup

Verification

Currently, the requirement to conduct a

verification of initial enrichment and burnup

of spent fuel that will be stored in Region 2

is included as a general requirement of TS

2.8. It is proposed to relocate this

requirement into a surveillance in TS 3.2,

Table 3-5, consistent with NUREG-1432.

The proposed changes for the surveillance

requirements consist of providing additional

testing requirements to ensure that the

Limiting Condition for Operations will be

met. One surveillance frequency related to

the sampling of the reactor coolant system

boron concentration during refueling

operations is being reduced from a frequency

of once per shift to once every 3 days.

However, this frequency is consistent with

the frequency of sampling during the

remainder of Mode 5 when fuel is in the

reactor and is more than adequate due to

the large volume (over 200,000 gallons) of

borated water required during refueling

operations. Therefore, the proposed changes

do not increase the probability or

consequences of an accident previously

evaluated.

ADMINISTRATIVE CHANGES

The remainder of TS 2.8 requirements of

refueling operations are proposed to be

reformatted into individual TS LCOs. It is

also proposed that sampling frequencies of

items contained in TS 3.2, Table 3-4, (page

3-19), be revised to incorporate frequencies

defined in TS 3.0.2. Therefore, frequencies

stated as once per 31 days will be noted as

‘‘M,’’ and frequencies stated as once per 7

days will be noted as ‘‘W.’’ These proposed

changes have no effect on 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 will be no physical alterations to the

plant configuration. No changes in operating

modes are proposed although minor changes

to the definitions of Cold Shutdown

Condition and Refueling Shutdown

Condition are proposed for clarification

purposes. The proposed changes incorporate

additional restrictions on the operation and

testing of equipment required to mitigate an

accident and to ensure the initial conditions

and assumptions of the design basis

accidents are maintained and controlled by

the Technical Specifications.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes ensure that

assumptions of the fuel handling accident are

maintained by Technical Specification

Limiting Condition for Operation and

surveillance requirements. The assumptions

of the fuel handling accident that may affect

a margin of safety are not being changed.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: February

25, 1997

Description of amendment request:

The proposed Technical Specifications

(TS) changes would amend the Limerick

Generating Station (LGS) Unit 1 and

Unit 2 Facility Operating Licenses

(FOLs), and Appendix B of the licenses

(i.e., Environmental Protection Plan

(EPP)), reflecting a corporate name

change from Philadelphia Electric

Company to PECO Energy Company. In

addition, the application would make

changes to the LGS Units 1 and 2, FOL,

and Appendix A (i.e., TS) of the

licenses, which would remove obsolete

information and correct typographical

errors.

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 TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The company name change and

typographical corrections are editorial and

will not alter the operation of equipment

assumed to be an initiator of any analyzed

event or transients previously evaluated. The

license provisions were satisfactorily

completed, and as such, have no effect on

any previously evaluated accident scenario.

The changes will not alter the operation of

equipment assumed to be available for the

mitigation of accidents or transients, nor will

they alter the operation of equipment

important to safety previously evaluated.

Therefore, the changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The company name change and

typographical corrections are editorial and

will not involve any physical changes to the

plant systems, structures, or components.

The license provisions were satisfactorily

completed, and as such, have no effect on

any previously evaluated accident scenario.

The proposed changes do not allow plant

operation in any mode that is not already

evaluated. The changes will not alter the

operation of equipment important to safety

previously evaluated.

Therefore, the changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The company name change and

typographical corrections are editorial and

will not affect the manner in which the

facility is operated, or change equipment or

features which affect the operational

characteristics of the facility. There is no

margin of safety as defined in the bases of

any TS regarding the name of the company,

or affected by the corrections or deletion of

obsolete license provisions.

Therefore, these 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30643

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: March

24, 1997

Description of amendment request:

The proposed Technical specifications

(TS) changes would delete the Drywell

and Suppression Chamber Purge System

operational time limit, and add a

surveillance requirement to ensure the

purge system large supply and exhaust

valves are closed as required.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

This activity does not increase the

probability of occurrence of an accident

previously evaluated in the SAR [Safety

Analysis Report]. This activity involves

deleting the allowable operating limit (180

hours each 365 days) for the Drywell and

Suppression Chamber Purge system, while

maintaining specific criteria for when the

valves are allowed to be open. These changes

do not increase the probability that this

system will be in service should a LOCA

[loss-of-coolant-accident] occur and does not

increase the probability that a LOCA will

occur. These changes also do not impact the

probability of occurrence of any anticipated

operational occurrence, other postulated

design basis accident, or other event in

which the plant was designed to respond.

This activity does not increase the

consequences of an accident previously

evaluated in the SAR. UFSAR [Updated Final

Safety Analysis Report] Section 9.4.5.1.2.2

for high volume purging, although limiting

the operating time the vent and purge system

is to be in service, evaluates the

consequences of a LOCA should the vent and

purge valves be open. System operating

procedures for venting and purging assure

the availability of SGTS [standby gas

treatment system] should a LOCA occur.

This activity will not increase the

probability of a LOCA occurring during the

time the Drywell and Suppression Chamber

Purge system is in operation as previously

evaluated. The Improved TS do not identify

a specific time limit value as long as the

valves are operated under the stated

conditions (inerting, de inerting, pressure

control, ALARA [as low as reasonably

achievable] or air quality considerations for

personnel entry or Surveillances that require

that the valves be open). These proposed

changes will incorporate the ITS [Improved

Technical Specifications] operational

controls which will result in the same order

of magnitude of equipment malfunction

probability as that provided by limiting

purging to 180 hours per 365 days. A LGS

[Limerick Generating Station] Level 2 PSA

[Probability Risk Assessment] Analysis was

performed to determine the additional risk

associated with changing the operating limit

from 90 hours to a nominal 500 hours each

365 days. This analysis concluded that the

increase in risk of containment failure is well

within the bounds of the EPRI [Electric

Power Research Institute] PSA Applications

Guideline for permanent changes and the

NRC [Nuclear Regulatory Commission] Staff’s

safety goal value of 1.0 E-6 per year of reactor

operation. Industry and LGS historical

operating experience confirms that the

purging lines are opened only for the

specified reasons stated in ITS and for

periods which do not exceed the current

magnitude of equipment malfunction

probability. Therefore, earlier engineering

judgment is being replaced by operating

experience.

Failure of the operating SGTS filter bank

following a LOCA has been found to be

acceptable due to the limited benefit derived

from SGTS for accident sequences important

to plant risk and the possibility that the

backup filter bank would be available.

Additionally, as discussed in UFSAR Section

9.4.5.1.2.2, the failure of SGTS during a

LOCA does not contribute to any significant

releases and is bounded by the analysis

performed to address containment

overpressure rupture.

Deleting the time limit restriction that the

vent and purge line isolation valves may be

open does not increase the probability that

these valves will not perform as designed

(close upon isolation signal) in response to a

LOCA. Removing the 180 hour requirement

will not increase the likelihood that the vent

and purge valves will be called upon to close

from that previously evaluated. UFSAR

Section 6.2 states that the containment purge

valves have undergone extensive testing and

analyses to demonstrate the operability of

these valves following a LOCA.

These changes do not directly or indirectly

degrade the performance of any other safety

system (assumed to function in the accident

analysis) design basis. The potential for other

equipment failures in the reactor enclosure

due to duct impact, impingement, and the

resulting environmental conditions was

previously evaluated in the LGS SAR. It was

concluded that the environmental

qualifications for the LGS equipment are

sufficient to ensure operability under the

predicted environmental condition, and, the

potential does not exist for impact or

impingement - related damage to essential

equipment. Maintaining the existing SAR

analysis and retaining operating criteria for

opening the containment purge valves,

demonstrates that the risk of equipment

failure and resulting radiological

consequences will not increase.

Therefore, deleting the TS operating limit

for the Drywell and Suppression Chamber

Purge system from 180 hours each 365 days

and the addition of a TS Surveillance

Requirement verifying that the purge valves

are closed under certain conditions does not

increase the probability or consequences of

an accident previously evaluated.

2. The proposed Technical Specifications

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

This activity does not change the function

of the Drywell and Suppression Chamber

Purge system, the containment isolation

system, or SGTS as previously evaluated.

Deleting the operational time limit that the

vent and purge system is in service and the

addition of a surveillance requirement does

not create an accident initiator not already

considered.

In addition, this activity does not create a

failure mode not considered. All evaluated

equipment failures that could occur as a

result of a LOCA during high volume purging

have previously been identified and

evaluated. Therefore, these changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed Technical Specifications

changes do not involve a significant

reduction in a margin of safety.

The bases of TS 3.6.1.8 state that the 180

hour each 365 day operating limit for the

Drywell and Suppression Chamber Purge

system is imposed to protect the integrity of

the SGTS filters. The LGS Offsite Dose

Calculation Manual assures the availability of

the backup SGTS filter train during operation

of the vent and purge system. Furthermore,

deleting the operating limit (180 hours each

365 days) does not reduce the margin of

safety since specific criteria for opening the

purge valves is being maintained and does

not involve an increase in risk. 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: April 9,

1997

Description of amendment request:

The proposed Technical Specifications

(TS) changes would clarify existing

battery specific gravity requirements,

delete the requirement to correct

specific gravity values based on

electrolyte level, and allow the use of

charging current measurements to verify

the battery—s state of charge.

Basis for proposed no significant

hazards consideration determination:
30644 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

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 Technical Specifications

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

Changes to Technical Specifications

surveillance requirements for specific gravity

and Technical Specifications Bases

commitments do not change the frequency or

consequences of any accident previously

evaluated. The proposed changes which

commit to IEEE Standard 450-1995 for

specific gravity testing, providing battery

charging current as an alternate method to

specific gravity measurements, and

eliminating the commitment to perform

electrolyte level correction do not prevent the

DC system from performing its intended

safety function. The proposed changes to the

Technical Specification battery surveillance

requirements and commitment to IEEE

Standard 450-1995 for specific gravity are in

accordance with current industry practices.

These changes do not reduce the readiness

and performance of the 1E DC power system

to perform its intended function during a

design basis event.

The proposed changes do not affect seismic

specifications, separation criteria or

environmental qualifications. The proposed

changes do not impose an increase in or more

severe test requirements, an increase in the

frequency of operation, reduce independence

or redundancy, modify the system or

equipment protective features, introduce new

equipment failures or impose additional

loads than any previously evaluated. The

Class 1E battery system will continue to meet

all of the design standards applicable to the

system and will not cause the system to

operate outside of its design or testing limits.

Batteries or battery chargers and their

failure are not initiators of the accidents

previously evaluated. The proposed changes

do not affect, degrade or prevent the response

of active or passive systems described or

assumed in the LGS accidents previously

evaluated. In addition, the proposed TS

changes will improve the availability of the

station batteries.

Therefore, the changes will not increase

the probability or consequences of an

accident previously evaluated.

2. The proposed Technical Specifications

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The proposed Technical Specifications

changes which will revise the surveillance

requirements and the TS Bases, do not

increase the failure rate of the battery. The

proposed changes clarify and enhance

Operation’s focus on the key battery

parameters which will improve the

availability of the station batteries. The

station batteries are not accident initiators.

The single failure of an electrical component

was previously evaluated in the LGS accident

analysis. Unexpected failures beyond the

postulated single failure are no more likely

to occur under the clarified surveillance

requirements.

Therefore, these changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed Technical Specifications

changes do not involve a significant

reduction in a margin of safety.

The revision clarifies and reduces the

battery surveillance requirements for specific

gravity. The revision eliminates the

possibility for misinterpretation and provides

consistency of the surveillance requirements.

The specific gravity value for each connected

cell is being revised to reflect a discrete

number which meets the existing

manufacturer’s recommendations and does

not differ from the value described in the

present bases. LGS is currently committed to

earlier revisions of IEEE Standard 450 (i.e.,

1975 and 1980), and the incorporation of

IEEE Standard 450-1995 for specific gravity

will reflect current industry practices

regarding specific gravity.

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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket No.

50-390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: April 30,

1997 (TS 97-01)

Description of amendment request:

The proposed amendment would

change the design features section of the

Technical Specifications to provide for

insertion of Lead Test Assemblies

(LTAs) containing Tritium Producing

Burnable Absorber Rods (TPBARs) in

the Watts Bar Nuclear Plant (WBN)

reactor core during Cycle 2. The

purpose of the change is to provide

irradiation services to support U.S.

Department of Energy (DOE)

investigations into the feasibility of

using commercial light water reactors to

maintain the DOE inventory of tritium.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

LTAs do not adversely affect reactor

neutronic or thermal-hydraulic performance;

therefore, they do not significantly increase

the probability of accidents or equipment

malfunctions while in the reactor. The

neutronic behavior of the LTAs mimics that

of standard burnable absorbers with only

slight differences which are accommodated

in the core design. The reload safety analysis

performed for WBN Unit 1, Cycle 2 will

confirm that any minor effects of LTAs on the

reload core will be within established fuel

design limits.

As described in DOE Technical Report

PNNL-11419, Revision 1, the LTA design is

robust to all accident conditions except the

large loss of coolant accident where the rods

are susceptible to failure. However, the

failure of the small number of TPBARS rods

has been determined to have an insignificant

effect on the thermal hydraulic response of

the core to this event.

The impacts of LTAs on the radiological

consequences for certain postulated events

[as shown in Table 6-1 of the licensee’s

submittal, including Large Break LOCAsb

are very small, and they remain within 10

CFR 100 regulatory limits. The additional

offsite doses due to tritium leakage from the

containment are small with respect to loss of

coolant accident source terms and are well

within regulatory limits.

The LTAs will not result in an increase in

combustible gas released to the containment.

Therefore, the LTAs do not result in a

significant increase in the consequences of

those previously considered.

Analysis has shown that TPBARs will not

fail during Condition I through IV events,

with the exception of a Large Break LOCA.

The radiological consequences of the nonLarge-Break LOCA events are essentially

unchanged by the expected TPBAR tritium

leakage to reactor coolant, and doses remain

within a small fraction of 10 CFR 100

regulatory limits. Therefore, there is no

significant increase in the consequences of

these previously evaluated accidents.

The expected occupational and offsite

doses, as reported in Technical report PNNL11419, Revision 1, resulting from release of

tritium from TPBARs over the plant

operating cycle, including refueling, are not

significantly increased and are within

applicable regulatory limits.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

LTAs have been designed to be compatible

with existing Westinghouse 17x17 fuel

assemblies and conventional Burnable

Poison Rod Assembly (BPRA) handling tools,

equipment, and procedures, and therefore no

new accidents or equipment malfunctions are

created by the handling of LTAs.

LTAs use materials with known and

predictable performance characteristics and

are compatible with PWR coolant. The LTA

design has specifically included material

similar to those used in standard burnable

absorber rods with the exception of internal
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30645

assemblies used in the production and

retention of tritium. As described in the

technical report, these materials are

compatible with the reactor coolant system

and the core design. For the irradiation

proposed, the quantities of these materials is

small. Therefore, no new accidents or

equipment malfunctions are created by the

presence of the LTAs in the reactor coolant

system.

Thermal-hydraulic criteria have been

established to ensure that TPBARs will not

fail during Condition I or II events. Analysis

has shown that TPBARs, appropriately

positioned in the core, operate within the

established thermal-hydraulic criteria.

Therefore, no new accidents or equipment

malfunctions are created by the presence of

the LTAs in the reactor.

Analysis has shown that TPBARs will not

fail during Condition III and IV events, with

the exception of a large-break loss-of-coolantaccident. The radiological consequences of

these events are small, with doses that are a

small fraction of the 10 CFR 100 limits.

Therefore there is no significant increase in

consequences of these previously evaluated

accidents.

LTAs do not adversely affect reactor

neutronic or thermal-hydraulic performance;

therefore, they do not create the possibility

of accidents or equipment malfunctions of a

different type than previously evaluated

while in the reactor.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

LTAs do not adversely affect reactor

neutronic or thermal-hydraulic performance.

Analysis indicates that reactor core behavior

and offsite doses remain relatively

unchanged. TPBAR performance under

Condition I, II, III, and IV events are very

similar to standard burnable absorber rods

previously evaluated. For these reasons, the

proposed amendment 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

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 amendment request: April 18,

1997

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section 3/

4.3.2, ‘‘Safety System Instrumentation,’’

and TS Section 3/4.5.2, ‘‘Emergency

Core Cooling Systems - ECCS

Subsystems - Tavg (greater than or equal

to) 280°F.’’ Certain surveillance

intervals would be changed from 18

months to once each refueling interval,

and certain setpoints would be changed.

The associated bases would also be

changed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The Davis-Besse Nuclear Power Station,

Unit No. 1 (DBNPS) has reviewed the

proposed changes and determined that a

significant hazards consideration does not

exist because operation of the DBNPS, in

accordance with these changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because the initiation of such

accidents are not affected by the proposed

revisions to increase the surveillance test

intervals from 18 to 24 months for TS 3/

4.3.2.1, ‘‘Safety Features Actuation System

Instrumentation,’’ and TS 3/4.5.2,

‘‘Emergency Core Cooling Systems - ECCS

Subsystems - Tavg (greater than or equal to)

280—F.’’ Initiating conditions and

assumptions remain as previously analyzed

for accidents in the DBNPS Updated Safety

Analysis Report.

Results of the instrument drift study

analysis and review of historical 18-month

surveillance data and applicable

maintenance records support an increase in

the surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because: the projected

instrument errors caused by drift are

bounded by the existing setpoint analysis or

a new analysis has been performed

incorporating a more conservative setpoint;

and no potential for a significant increase in

a failure rate of a system or component was

identified during surveillance data and

applicable maintenance records reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

The proposed revisions to Allowable

Values for Safety Features Actuation System

(SFAS) Reactor Coolant System (RCS)

Pressure - Low, RCS Pressure - Low-Low,

RCS Pressure - Low-Low bypass permissive,

and Decay Heat Isolation Valve and

Pressurizer Heater Interlocks have no bearing

on the probability of the initiation of an

accident previously evaluated.

The application of the Allowable Value to

only the Channel Functional Test and not the

Channel Calibration, the proposed deletion of

the Trip Setpoints, the proposed revision of

the TS 3.3.2.1 Limiting Condition for

Operation (LCO) and Action Statement

3.3.2.1.a, and the proposed revisions to

Actions 13 and 14 of TS Table 3.3-3, are

associated with the proposed revision of the

Allowable Values for SFAS RCS Pressure -

Low, RCS Pressure - Low-Low, and Decay

Heat Isolation Valve and Pressurizer Heater

Interlocks, and are consistent with NUREG1430, Revision 1, ‘‘Standard Technical

Specifications, Babcock and Wilcox Plants,’’

dated April 1995. The proposed revisions

have no bearing on the probability of the

initiation of an accident previously

evaluated.

The proposed changes to TS Bases 3/4.3.1

and 3/4.3.2, ‘‘Reactor Protection System and

Safety System Instrumentation,’’ and TS

Bases 3/4.5.2 and 3/4.5.3, ‘‘ECCS

Subsystems,’’ are administrative changes

associated with the other proposed changes,

and do not affect previously analyzed

accidents.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the slight increase in doses

due to a letdown line break event as a result

of the proposed change to the SFAS RCS

Pressure - Low Allowable Value still satisfy

the NRC Standard Review Plan Section

15.6.2 acceptance criteria that doses do not

exceed a small fraction (10%) of the 10 CFR

100 guideline values. The remaining

proposed changes to Allowable Values, and

the other changes proposed by this License

Amendment Request do not increase the

radiological consequences of previously

analyzed accidents because the source term,

containment isolation, or radiological

releases are not being changed by the

proposed revisions.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated, for the reasons

discussed below.

No changes are being proposed to the type

of testing currently being performed, only to

the length of the surveillance test interval.

Results of the instrument drift study

analysis and review of historical 18-month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because: the

projected instrument errors caused by drift

are bounded by the existing setpoint analysis

or a new analysis has been performed

incorporating a more conservative setpoint;

and no potential for a significant increase in

a failure rate of a system or component was

identified during surveillance data and

applicable maintenance records reviews.

The proposed revisions to Allowable

Values for SFAS RCS Pressure - Low, RCS

Pressure - Low-Low, RCS Pressure Low-Low

bypass permissive, and Decay Heat Isolation

Valve and Pressurizer Heater Interlocks, do

not alter the type of any testing currently

being performed.

The application of the Allowable Value to

only the Channel Functional Test and not the

Channel Calibration, the proposed deletion of

the Trip Setpoints, revision of the TS 3.3.2.1

LCO and Action Statement 3.3.2.1.a, and the

proposed revisions to Actions 13 and 14 of
30646 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

TS Table 3.3-3, are associated with the

proposed revision to the Allowable Values

for SFAS RCS Pressure - Low, RCS Pressure

- Low-Low, RCS Pressure Low-Low bypass

permissive, and Decay Heat Isolation Valve

and Pressurizer Heater Interlocks, and are

consistent with NUREG-1430, Revision 1,

‘‘Standard Technical Specifications, Babcock

and Wilcox Plants,’’ dated April 1995. The

proposed revisions do not alter the type of

testing currently being performed.

The proposed changes to TS Bases 3/4.3.1

and 3/4.3.2, ‘‘Reactor Protection System and

Safety System Instrumentation,’’ and TS

Bases 3/4.5.2 and 3/4.5.3, ‘‘ECCS

Subsystems,’’ are administrative changes

associated with the other proposed changes,

and do not alter any testing currently being

performed.

3. Not involve a significant reduction in a

margin of safety. The results of the

instrument drift study analysis and review of

historical 18-month surveillance data and

applicable maintenance records support an

increase in the surveillance test intervals

from 18 to 24 months (and up to 30 months

on a non-routine basis) because: the projected

instrument errors caused by drift are

bounded by the existing setpoint analysis or

a new analysis has been performed

incorporating a more conservative setpoint;

and no potential for a significant increase in

a failure rate of a system or component was

identified during surveillance data and

applicable maintenance records reviews.

Existing system and component redundancy

is not affected by these proposed changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences, consequently there

are no significant reductions in a margin of

safety.

The NRC staff has reviewed the

licensees’ 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 Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Attorney for licensees: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

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 amendment request: April 18,

1997

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section 3/

4.7.6, ‘‘Plant Systems - Control Room

Emergency Ventilation System.’’

Additional Limiting Conditions for

Operation (LCO) would be added

related to the availability of the station

vent normal range radiation monitoring

instrumentation. The associated TS

bases would also be modified consistent

with these changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The Davis-Besse Nuclear Power Station has

reviewed the proposed changes and

determined that a significant hazards

consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station (DBNPS), Unit No. 1, in accordance

with this change would not:

1a. Involve a significant increase in the

probability of an accident previously

evaluated because no accident initiators,

conditions, or assumptions are affected by

the proposed changes.

The proposed change to LCO 3.7.6.1 would

include new required Action statements in

the event that one or both channels of station

vent normal range radiation monitoring

instrumentation become inoperable. In the

event that one channel is inoperable for

greater than 7 days, or in the event that both

channels are inoperable, the proposed Action

statement would require that the control

room normal ventilation system be isolated

and at least one Control Room Emergency

Ventilation System (CREVS) train be placed

in operation.

Under the proposed actions, the ventilation

systems would be placed in a state equivalent

to that which occurs were a high radiation

isolation to occur. These proposed changes

have no bearing on the probability of an

accident.

The proposed change to the terminology

utilized in Surveillance Requirement (SR)

4.7.6.1.e is an administrative change made to

make the terminology consistent with the

proposed new Action statements. The

proposed changes to Bases 3/4.7.6 are

administrative changes consistent with the

proposed changes to LCO 3.7.6.1. These

changes have no bearing on the probability

of an accident.

1b. Involve a significant increase in the

consequences of an accident previously

evaluated because the proposed changes do

not change the source term, containment

isolation, or allowable releases.

As described above, under the proposed

new LCO 3.7.6.1 Actions, in the event that

one station vent normal range radiation

monitoring instrumentation channel is

inoperable for greater than 7 days, or in the

event that both channels are inoperable, the

ventilation systems would be placed in a

state equivalent to that which occurs were a

high radiation isolation to occur. Therefore,

in the unlikely event of an accident requiring

control room isolation while in this

condition, the dose consequences to control

room operators would be unchanged.

The proposed change to the terminology

utilized in SR 4.7.6.1.e is an administrative

change made to make the terminology

consistent with the proposed new Action

statements. The proposed changes to Bases 3/

4.7.6 are administrative changes consistent

with the proposed changes to LCO 3.7.6.1.

These changes have no bearing on the

consequences of an accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new

accident initiators or assumptions are

introduced by the proposed changes.

As described above, under the proposed

new LCO 3.7.6.1 Actions, in the event that

one station vent normal range radiation

monitoring instrumentation channel is

inoperable for greater than 7 days, or in the

event that both channels are inoperable,

theventilation systems would be placed in a

state equivalent to that which occurs were a

high radiation isolation to occur. Operation

of the equipment and components in this

manner would not introduce the possibility

of any new or different kinds of accidents.

The proposed change to the terminology

utilized in SR 4.7.6.1.e is an administrative

change made to make the terminology

consistent with the proposed new Action

statements. The proposed changes to Bases 3/

4.7.6 are administrative changes consistent

with the proposed changes to LCO 3.7.6.1.

These changes would not introduce the

possibility of any new or different kinds of

accidents.

3. Involve a significant reduction in a

margin of safety because the proposed

changes to the Action under LCO 3.7.6.1

ensure that control room isolation capability

is maintained in the event a station vent

radiation monitor is inoperable. The

proposed allowable outage time of 7 days for

one inoperable channel is consistent with the

presently allowable outage time for one

inoperable CREVS. The proposed Action to

place at least one CREVS train in operation

within 1 hour, in the event both channels of

radiation monitoring become inoperable, is

more conservative than the present Action

which requires that a plant shutdown

commence within 1 hour, but does not

require the CREVS be placed in operation.

The proposed change to the terminology

utilized in SR 4.7.6.1.e is an administrative

change made to make the terminology

consistent with the proposed new Action

statements. The proposed changes to Bases 3/

4.7.6 are administrative changes consistent

with the proposed changes to LCO 3.7.6.1.

These changes would not affect the margin of

safety.The NRC staff has reviewed the

licensees’ 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 Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Basis for proposed no significant

hazards consideration determination:
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30647

Attorney for licensees: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: August

22, 1996

Description of amendment request:

The proposed change would remove the

action statement of Technical

Specification (TS) Section 3.2.G, Table

3.2.6, Note 7, requiring reactor

shutdown after 30 days of inoperability

of the high range stack gas monitor and

substitute an action statement consistent

with the guidance provided in NRC

Generic Letter 83-36.

The high range stack monitor

provides an estimate of gross stack

activity that has exceeded the upper

limit of the normal range

instrumentation. The high range

monitor reading serves as input to dose

projection systems for initial estimation

of off-site conditions. The monitor

reading would be used prior to the

acquisition of stack isotopic sample data

which would provide a more accurate

indication of stack activity.

The licensee stated, among other

things, that due to the passivefunction

of the instruments and the ability to

monitor this parameter utilizing

alternate methods, it is not appropriate

to impose stringentrequirements on the

operation of the unit. This monitor is

identified in the Vermont Yankee

Regulatory Guide 1.97 submittal as

Category 2, Type E. This monitor

provides post-accident information for

use in determining the magnitude of the

release of radioactive materials and for

monitoring such release. However, the

high range stack monitor does not have

any safety function associated with the

prevention or automatic mitigation of

design-basis accidents, neither does it

provide primary information needed to

permit the control room operating

personnel to take required manually

controlled actions.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91 (a),the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below.

[(1) The proposed TS change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.]

The High Range Stack Monitor is a RG

[Regulatory Guide] 1.97, Category 2, Type E

instrument with no specified safety function

associated with the prevention or automatic

mitigation of design basis accidents, neither

does it provide primary information needed

to permit the control room operating

personnel to take required manually

controlled actions. The proposed change to

the action statement associated with this

monitor will not change the function of this

monitor, and since the monitor is not

assumed to initiate any accidents, nor

function to mitigate any accidents, this

change will not significantly increase the

probability or consequences of any

previously analyzed accident.

[(2) The proposed TS change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.]

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change will

still ensure effective methods are available to

assess post accident conditions. Thus, this

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

[(3) The proposed TS change does not

involve a significant reduction in a margin of

safety.]

The proposed change to the action

statement associated with this monitor will

not change the function of this monitor, and

since the monitor is not assumed to function

for the prevention or mitigation of any

previously evaluated accidents, this change

will not significantly 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Attorney for licensee: R. K. Gad, III,

Ropes and Gray, One International

Place, Boston, MA 02110-2624

NRC Project Director: Patrick D.

Milano, Acting

Wisconsin Electric Power Company,

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

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: January

24, 1997, as supplemented on May 15,

1997 (TSCR 193)

Description of amendment request:

The proposed amendments (Point Beach

Nuclear Plant (PBNP) Technical

Specifications (TS) Change Request

(TSCR) 193) would revise TS 15.5.4,

‘‘Fuel Storage,’’ to increase fuel

assembly enrichment limits to 5.0 w/o

U-235 while maintaining Keff in the

storage pools (spent fuel pool and new

fuel storage racks) less than 0.95. The

May 15, 1997, supplement provided a

revised no significant hazards

consideration determination that

superseded the licensee’s determination

noticed in the Federal Register on April

23, 1997 (62 FR 19837).

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. Operation of this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The proposed changes do not involve a

change to structures, systems, or components

that would affect the probability or

consequences of an accident previously

evaluated in the PBNP Final Safety Analysis

Report (FSAR). The only relevant concern

with respect to increasing enrichment limits

in the spent fuel pool and new fuel storage

racks is one of criticality. The proposed

changes use the same criticality limit used in

the current Technical Specifications.

Therefore, margin to safe operation of Units

1 and 2 is maintained. The probability and

consequences of an accident previously

evaluated are dependent on this criticality

limit. Because the limit will not change, the

probability and consequences of those

accidents previously evaluated will not

change.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not involve a

change to the physical structure of the spent

fuel pool or of the plant. The proposed

increase in spent fuel pool and new fuel

storage racks fuel assembly enrichment limits

maintains the margin to safe operation of

Units 1 and 2 because the criticality limit for

the spent fuel pool and new fuel storage

racks will not change. The enrichment

increase does not affect any of the parameters

or conditions that contribute to the initiation

of any accidents. Because the criticality limit

remains the same, these changes have no

effect on plant operation or on the initiation

of any accidents. Therefore, the proposed

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The proposed changes maintain the margin

to safe operation of Units 1 and 2. The

margin of safety is based on the criticality

limit of the spent fuel pool and the new fuel

storage racks. Because this limit will not

change, the margin of safety will not be

affected. Therefore, the proposed changes

will not create a significant reduction in a

margin of safety.
30648 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John N. Hannon

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: April 23,

1997

Description of amendment request:

This request proposes to revise

Technical Specification 3/4.9.4,

Containment Building Penetrations, and

its associated Bases section, to allow

selected containment isolation valves to

be opened under administrative controls

during periods of core alterations or

movement of irradiated fuel inside

containment.

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 involves changes to

the Technical Specification requirements for

containment closure which is an accident

mitigating feature. The changes would not

affect the likelihood of occurrence of any

accidents previously evaluated. The

proposed change does not involve any

hardware or plant design changes. The

containment leakage value is not assumed to

be an initiator of any analyzed event.

Containment isolation valves and temporary

closure devices serve to limit the radiological

consequences of accidents. The proposed

change would ensure the service air and

breathing air manual isolation valves will

perform their required containment closure

function and will serve to limit the

consequences of a fuel handling accident as

described in the USAR, such that the results

of the analyses in the USAR remain

bounding. In considering the consequences

of a design basis fuel handling accident

inside containment, the assumptions in the

analysis take no credit for the containment as

a barrier to prevent the postulated release of

radioactivity. For events that could occur

during CORE ALTERATIONS or movement

of irradiated fuel assemblies, containment

closure is considered a defense-in-depth

boundary to prevent uncontrolled release of

radioactivity. Additionally, the proposed

change does not impose any new safety

analyses limits or alter the plant’s ability to

detect and mitigate events. Therefore, this

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.

The proposed change involves reliance on

manual actuation of containment penetration

valves (Service Air valves KA V-039 and KA

V-118 and Breathing Air valves KB V-001

and KB V-002 are manual valves) to block the

unimpeded flow of the containment

atmosphere to the environment under certain

conditions. The proposed change would not

necessitate a physical alteration of the plant

features that provide core cooling or

subcriticality (no new or different type of

equipment will be installed) or changes in

parameters governing plant operation during

CORE ALTERATIONS or movement of

irradiated fuel in containment. Thus, this

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 is similar to the use

of administrative controls to isolate an open

containment airlock door. The use of

administrative controls in this manner has

been approved by the NRC (WCGS Technical

Specification Amendment 95) for plant

operations that would not require the

containment to maintain a pressure

boundary. This scenario is applicable during

plant shutdown for refueling when CORE

ALTERATIONS and movement of irradiated

fuel assemblies in the containment occur.

Accidental damage to spent fuel during these

operations is classified as a fuel handling

accident. The proposed change has been

developed considering the importance of the

containment boundary in limiting the

consequences of a design basis fuel handling

accident. The proposed change allows for

protection equivalent to that provided by

previously approved methods of containment

closure. Considering the probability of an

event that would challenge the containment

boundary, the alternative protection provided

by this change, and the operational

requirements to occasionally open these

penetrations, the proposed change is

acceptable and any reduction in the margin

of safety is insignificant.

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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, N.W., Washington, D.C.

20037

NRC Project Director: William H.

Bateman

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.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of application for amendment:

January 24, 1997, as supplemented

March 27, 1997

Brief description of amendment: The

proposed amendment will update the
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30649

Safety Limit Minimum Critical Power

Ratio (SLMCPR) in Technical

Specification 2.1.2 and the associated

Bases section to reflect the results of the

latest cycle-specific calculation

performed for the Pilgrim Nuclear

Power Station Operating Cycle 12. In

addition, the values provided in Note 5

of Table 3.2.C.1, which are based on the

SLMCPR values, have been revised as a

result of the changes to the SLMCPR

value.

Date of issuance: April 7, 1997

Effective date: April 7, 1997

Amendment No.: 171

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6568) The March 27, 1997,

supplemental letter provided clarifying

information that did not change the

initial proposed no significant hazards

consideration. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 7, 1997 No significant hazards

consideration comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Commonwealth Edison Company,

Docket Nos. STN 50-456 and STN 50-

457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of application for amendments:

August 19, 1996, as supplemented on

February 5, March 13, April 29 and

April 30, 1997.

Brief description of amendments: The

amendment would revise Technical

Specification (TS) Section 4.4.5.2 to

extend, for one additional operating

cycle (i.e., Cycle 7), the 1.0 volt and 3.0

volt interim plugging criteria (IPC)

which were added to the Braidwood,

Unit 1, TSs by License Amendment No.

69, issued on November 9, 1995.

Additionally, this amendment to the

Braidwood, Unit 1, license added some

definitions and reporting requirements

to TS Section 4.4.5.2 and modified the

designations for the IPC models in TS

Bases Section 3/4.4.4.5. Braidwood,

Unit 1, Cycle 7, will end in fall 1998.

While there are no revisions to the TS

for Braidwood, Unit 2, both units are

being amended to maintain the

continuity of the amendment numbers.

Date of issuance: May 14, 1997.

Date of effective: Immediately, to be

implemented within 30 days.

Amendment Nos.: 82

Facility Operating License Nos. NPF72 and NPF-77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6570). The February 5, March 13, April

29 and April 30, 1997, submittals

provided clarifying technical

information that did not affect the initial

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated May 14, 1997. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wilmington Public Library,

201 S. Kankakee Street, Wilmington,

Illinois 60481.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

June 20, 1996, as supplemented

December 30, 1996, and March 5, 1997.

Brief description of amendments: The

amendments would change the TSs by

incorporating an NRC-approved thermal

limit licensing methodology in the list

of approved methodologies used in

establishing the fuel cycle-specific

thermal limits. In addition, the

proposed amendments would change

the TSs to reflect the use of Siemens

Power Corporation (SPC) ATRIUM-9B

fuel for the first time at Dresden, Units

2 or 3. The proposed amendments

would also correct minor editorial items

in the TSs.

In March 1997, the NRC staff

performed an audit of the application of

Advanced Nuclear Fuel for Boiling

Water Reactors (ANFB) to ATRIUM-9

fuel. The staff raised concerns

associated with the ATRIUM-9B fuel

additive constant uncertainty used as

input to the NRC-approved

methodology for the calculation of

minimum critical power ratio (MCPR).

In response to the audit findings, by

letter dated April 18, 1997, SPC

submitted a generic topical report (ANF1125(P) Supplement 1 Appendix D),

which is currently under staff review,

for the future reload analysis in the

safety limit MCPR calculation. The staff

schedule for the review of the topical

report will not be timely enough for the

resolution of the ATRIUM-9B MCPR

issue to support reload and restart of

Dresden, Unit 3. Therefore, by letters

dated May 2 and May 6, 1997, ComEd

provided additional information

concerning the MCPR issues and how it

will affect the Dresden, Unit 3, D3R15

fuel cycle and provided additional

information concerning the ATRIUM-9B

fuel design and shutdown margin that

are applicable during refueling and

shutdown.

The staff is currently reviewing the

licensee’s May 2 and May 6, 1997,

letters. To be more timely and support

the reload schedule for Dresden, Unit 3

(currently scheduled for May 20, 1997),

the staff has chosen to split its

consideration of the proposed

amendments into two parts. The first

part of the amendment package now

being evaluated would modify Section

5.3.A, ‘‘Design Features’’ of the TSs to

reflect use of the ATRIUM-9B fuel

design and would include two SPC

topical reports in TS Section 6.9.A.6,

‘‘Core Operating Limits Report,’’ to

reflect mechanical design criteria for

this fuel. This change would allow this

fuel to be loaded into the core only

under Operational Modes 3 (Hot

Shutdown), 4 (Cold Shutdown), and 5

(Refueling) and does not permit startup

or power operation using the ATRIUM9B fuel.

Date of issuance: May 16, 1997

Date of effective: Immediately, to be

implemented within 30 days.

Amendment Nos.: 159 and 154

Facility Operating License Nos. DPR19 and DPR-25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17227).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 16, 1997 No

significant hazards consideration

comments received: No

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Commonwealth Edison Company,

Docket No. 50-265, Quad Cities Nuclear

Power Station, Unit 2, Rock Island

County, Illinois

Date of application for amendment:

April 21, 1997

Brief description of amendment: The

amendment increases the minimum

critical power ratio safety limit for Unit

2 and adds a Siemens Power

Corporation reference to the Technical

Specifications (TS) to allow plant

operation in Operational Modes 1 and 2.

Date of issuance: May 22, 1997

Date of effective: Immediately, to be

implemented within 30 days.

Amendment No.: 174

Facility Operating License No. DPR30: The amendment revised the TSs.

Public comments requested as to

proposed no significant hazards

consideration: Yes (62 FR 23499 dated

April 30, 1997). This notice provided an

opportunity to submit comments on the

Commission’s proposed no significant
30650 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by May 30, 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 amendment. The Commission’s

related evaluation of the amendment,

finding of exigent circumstances, and

final no significant hazards

consideration determination are

contained in a Safety Evaluation dated

May 22, 1997.

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan Date of application for

amendment: December 2, 1996 (NRC96-0134)

Brief description of amendment: The

amendment revises TS 3.1.4.3, TS Table

3.3.6-1, and TS Table 4.3.6-1 to change

the operability requirements for the Rod

Block Monitor (RBM). Specifically, the

revision requires the RBM to be

operable when reactor thermal power is

greater than or equal to 30 percent of

rated thermal power.

Date of issuance: May 15, 1997

Date of effective: May 15, 1997, with

full implementation within 60 days

Amendment No.: 112

Facility Operating License No. NPF43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 124)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 15, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

Duke Power Company, Docket Nos. 50-

269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

April 29, 1997

Brief description of amendments: The

amendments incorporate a license

condition that will allow revisions to

the Oconee Updated Final Safety

Analysis Report (UFSAR) that clarifies

the main turbine-generated missile

protection criteria.

Date of issuance: May 16, 1997

Date of effective: As of the date of

issuance and implementation is the

incorporation in the UFSAR the changes

described in Duke Power Company’s

application dated April 29, 1997

Amendment Nos.: 224, 224, and 221

Facility Operating License Nos. DPR38, DPR-47, and DPR-55: The

amendments revised the UFSAR and

added a new License Condition. Public

comments requested as to proposed no

significant hazards consideration: Yes.

(62 FR 24512 dated May 5, 1997). The

notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received as of the

date of issuance. The notice also

provided for an opportunity to request

a hearing by June 9, 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 final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated May 16, 1997.

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

November 26, 1996, as supplemented

February 12, 1997.

Brief description of amendment: The

amendment changes the allowable

primary-to-secondary leak rate and in

the Surveillance Requirements section

of the TSs it changes the acceptance

criteria for steam generator tubes. The

amendment changes the reference that

is included in the tube acceptance

criteria from Combustion Engineering

topical report CEN-601-P Revision 01-P

to CEN-630-P, Revision 01.

Date of issuance: May 20, 1997

Date of effective: May 20, 1997, to be

implemented within 30 days.

Amendment No.: 184

Facility Operating License No. NPF-6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64376) The February 12, 1997, submittal

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

May 20, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

April 4, 1995, as supplemented by

letters dated August 25, 1995, and April

18, 1997.

Brief description of amendment: The

amendment changes the required

frequency for inspecting reactor coolant

pump flywheels.

Date of issuance: May 20, 1997

Date of effective: May 20, 1997, to be

implemented within 30 days.

Amendment No.: 185

Facility Operating License No. NPF-6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 5, 1995, (60 FR 35069)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 20, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One,Unit No.

2, Pope County, Arkansas

Date of application for amendment:

October 7, 1996, as supplemented

February 10, and May 8, 1997

Brief description of amendment: The

amendment changes the channel

functional testing frequency for most of

the Reactor Protection System (RPS) and

Engineered Safety Feature Actuation

System (ESFAS) instrumentation from

monthly to every four months. In

addition, the amendment allows the use

of Cycle Independent Shape Annealing

Matrix (CISAM) methodology in the

Core Protection Calculators (CPCs).

Finally, the amendment makes a

number of administrative changes to the

Technical Specifications (TS) to clarify

the existing TS or correct previous

errors in the TS.

Date of issuance: May 21, 1997

Date of effective: May 21, 1997

Amendment No.: 186

Facility Operating License No. NPF-6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4346)

The Commission’s related evaluation of

the amendment is contained in a Safety
Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices 30651

Evaluation dated May 21, 1997 No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: March

27, 1997

Brief description of amendment: The

amendment changes TSs surveillance

requirements 4.5.2.d.3 and 4.5.2.d.4 by

increasing the required amount of

trisodium phosphate dodecahydrate

(TSP) stored in the containment sump

from 97.5 cubic feet to 380 cubic feet,

and adjusts the TSP sampling

requirement accordingly.

Date of issuance: May 15, 1997

Date of effective: May 15, 1997, to be

implemented within 60 days.

Amendment No.: 127

Facility Operating License No. NPF38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17234)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 15, 1997 No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: August

21, 1996, as supplemented by letter

dated March 17, 1997

Brief description of amendment: The

amendment approves revision of

Attachment 1 to the operating license

concerning design and testing

modifications in the Containment

Vacuum Relief System (CVR) that

penetrate the primary containment at

Waterford Steam Electric Station, Unit

3. The penetrations affected are

commonly referred to as Penetrations 53

and 65.

Date of issuance: May 20, 1997

Date of effective: May 20, 1997, to be

implemented within 90 days.

Amendment No.: 128

Facility Operating License No. NPF38: Amendment revised the Operating

License.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57484) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

May 20, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

September 13, 1996, as supplemented

by letter dated January 15, 1997.

Brief description of amendment: The

amendment revised the Technical

Specifications to permit the use of 10

CFR Part 50, Appendix J, Option B,

performance-based containment leakage

rate testing.

Date of issuance: May 19, 1997

Date of effective: May 19, 1997, to be

implemented within 60 days of the date

of issuance.

Amendment No.: 158

Facility Operating License No. DPR36: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57487) The January 15, 1997,

supplemental letter provided additional

clarifying information and did not

change the initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated May 19, 1997. No

significant hazards consideration

comments received: Yes. Comments

were submitted by Patrick J. Dostie on

behalf of the State of Maine by letter

dated April 15, 1997. The staff

responded by letter dated May 19, 1997.

Local Public Document Room

location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME

04578.

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: February

18, 1997, as supplemented by letter

dated February 26, 1997.

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications relating to the

reactor core fuel assembly design

features requirements contained in

Technical Specification 5.3.1, Fuel

Assemblies. The changes made by this

amendment allow for the limited

replacement of failed or damaged fuel

rods in fuel assemblies with solid

stainless steel or zirconium alloy filler

rods.

Date of issuance: May 13, 1997

Date of effective: May 13, 1997

Amendment No.: 51

Facility Operating License No. NPF86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 12, 1997 (62 FR 11496)

The licensee’s letter dated February 26,

1997, provided a correction to a

typographical error in the original

application but does 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 May 13, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833

Portland General Electric Company, et

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

Plant, Columbia County, Oregon

Date of application for amendment:

November 2, 1995.

Brief description of amendment: This

amendment changes the TS to reflect

changes in the organization as they

apply to oversite and management of the

Trojan Nuclear Plant.

Date of issuance: October 31, 1996

Date of effective: October 31, 1996

Amendment No.: 195

Facility Operating License No. NPF-1:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 27, 1995 (60 FR

58404) No significant hazards

consideration comments received: No.

Local Public Document Room

location: Branford Price Millar Library,

Portland State University, 934 S.W.

Harrison Street, P.O. Box 1151,

Portland, Oregon 97207

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

January 7, 1997

Brief description of amendments:

These amendments revise Technical

Specification (TS) 3/4.2.5 to incorporate

an exception to the provisions of TS

4.0.4 and to clarify the time at which the

surveillance can be performed by

adding that the surveillance is to be

performed within 24 hours after

attaining steady state conditions at or

above 90% rated thermal power. The

revised surveillance contains editorial

enhancements that clarify the
30652 Federal Register / Vol. 62, No. 107 / Wednesday, June 4, 1997 / Notices

surveillance requirement. Salem Unit 1

TS Table 3.2-1 is also being revised to

delete reference to three loop operation.

Date of issuance: May 8, 1997

Date of effective: Both units, as of date

of issuance, to be implemented prior to

entry into Mode 1 from the current

outage. Amendment Nos. 193 and 176

Facility Operating License Nos. DPR70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4353)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 8, 1997. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Southern Nuclear Operating Company,

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

Farley Nuclear Plant, Unit 1, Houston

County, Alabama

Date of amendment request: March

25, 1997

Brief description of amendments: The

amendment changes Technical

Specification 3/4.4.9, ‘‘Specific

Activity,’’ and the associated Bases to

reduce the limit associated with dose

equivalent iodine-131. The steady-state

dose equivalent iodine-131 limit would

be reduced by 40 percent from 0.5

[micro]Ci/gram to 0.3 [micro]Ci/gram

and the maximum instantaneous value

would be reduced by 40 percent from 30

[micro]Ci/gram to 18 [micro]Ci/gram.

Date of issuance: May 19, 1997

Date of effective: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 128

Facility Operating License Nos. NPF2 and NPF-8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: April 4, 1997 (62 FR 16201)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 19, 1997. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Dated at Rockville, Maryland, this 28th day

of May, 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

Director, Division of Reactor Projects III/IV,

Office of Nuclear Reactor Regulation

[Doc. 97–14395 Filed 6–3–97; 8:45 am]

BILLING CODE 7590–01–F

SECURITIES AND EXCHANGE

COMMISSION

[Rel. No. IC–22686; 811–4068]

Pacifica Funds Trust; Notice of

Application

May 28, 1997.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’).

ACTION: Notice of Application for

Deregistration under the Investment

Company Act of 1940 (the ‘‘Act’’).

APPLICANT: Pacifica Funds Trust.

RELEVANT ACT SECTION: Section 8(f).

SUMMARY OF APPLICATION: Applicant

seeks an order declaring that it has

ceased to be an investment company.

FILING DATES: The application was filed

on January 31, 1997, and amended on

May 9, 1997.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving applicant with a

copy of the request, personally or by

mail. Hearing requests should be

received by the SEC by 5:30 p.m. on

June 23, 1997, and should be

accompanied by proof of service on

applicant, in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.

Persons may request notification of a

hearing by writing to the SEC’s

Secretary.

ADDRESSES: Secretary, SEC, 450 5th

Street, N.W., Washington, D.C. 20549.

Applicant, 237 Park Avenue, Suite 910,

New York, NY 10017.

FOR FURTHER INFORMATION CONTACT:

Deepak T. Pai, Staff Attorney, at (202)

942–0574, or H.R. Hallock, Jr., Special

Counsel, at (202) 942–0564 (Division of

Investment Management, Office of

Investment Company Regulation).

SUPPLEMENTARY INFORMATION: The

following is a summary of the

application. The complete application

may be obtained for a fee at the SEC’s

Public Reference Branch.

Applicant’s Representations

1. Applicant is an open-end

management investment company

organized as a Massachusetts business

trust. On July 16, 1984, applicant

registered under the Act and filed a

registration statement on Form N–1A

pursuant to section 8(b) of the Act. The

registration statement became effective

on November 30, 1984. Applicant

commenced an initial public offering of

the first of its 23 series on December 26,

1985, and commenced its last initial

public offering of a series on November

15, 1995. Shares of five series were

never offered to the public.

2. First Interstate Capital

Management, Inc., served as applicant’s

investment adviser prior to April 1,

1996, when its parent company, First

Interstate Bancorp, merged into Wells

Fargo & Company. At a meeting on May

17, 1996, applicant’s board of trustees,

including a majority of the trustees who

are not ‘‘interested persons’’ of

applicant, approved entry into an

Agreement and Plan of Reorganization

(the ‘‘Reorganization Agreement’’) by

and between applicant and Stagecoach

Funds, Inc. (‘‘Stagecoach’’), an open-end

investment company advised by Wells

Fargo Bank, N.A. In reviewing the

proposed reorganization, applicant’s

board considered the potential impact of

the reorganization on applicant’s

shareholders, including (a) provisions

intended to avoid the dilution of

shareholder interests; (b) the

capabilities, practices, and resources of

the organizations that provided

investment advisory and certain other

services to applicant and Stagecoach; (c)

the shareholder services provided to

applicant’s shareholders, compared

with the shareholder services provided

to Stagecoach shareholders; (d) the

investment objectives, policies and

limitations of each series of applicant

and the corresponding series of

Stagecoach; (e) the historical investment

performance of each series of applicant

and the corresponding series of

Stagecoach; (f) the historical and

projected operating expenses of each

series of applicant and the

corresponding series of Stagecoach; and

(g) the anticipated tax consequences of

the reorganization.

3. Based upon its evaluation of the

information presented, applicant’s

board of trustees determined that the

reorganization was in the best interests

of the shareholders of each series of

applicant, and that the interests of the

shareholders of each series would not be

diluted. An amendment to the

Reorganization Agreement was

subsequently approved by the

applicant’s board of trustees on August

15, 1996, which provided that, because

of tax considerations, certain liabilities

of one of applicant’s 23 series (Pacifica

Asset Preservation Fund) would be

retained by that series rather than

transferred to its corresponding series of

Stagecoach.

4. On or about June 6, 1996, proxy

materials for a special shareholders

meeting were distributed to applicant’s