Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25101

Safeguards Policy and Procedures Letter

1–50, Revision 1, is not warranted.

Alternatives to the Proposed Action

The proposed action is to amend NRC

Source Material License SUA–648, for

reclamation of the Heap Leach Area, as

requested by Umetco. Therefore, the

principal alternatives available to NRC

are to:

1. Approve the license amendment

request as submitted; or

2. Amend the license with such

additional conditions as are considered

necessary or appropriate to protect

public health and safety and the

environment; or

3. Deny the amendment request.

Based on its review, the NRC staff has

concluded that the environmental

impacts associated with the proposed

action do not warrant either the limiting

of Umetco’s future operations or the

denial of the license amendment.

Additionally, in the TER prepared for

this action, the staff has reviewed the

licensee’s proposed action with respect

to the criteria for reclamation, specified

in 10 CFR Part 40, Appendix A, and has

no basis for denial of the proposed

action. Therefore, the staff considers

that Alternative 1 is the appropriate

alternative for selection.

Finding of No Significant Impact

The NRC staff has prepared an EA for

the proposed renewal of NRC Source

Material License SUA–648. On the basis

of this assessment, the NRC staff has

concluded that the environmental

impacts that may result from the

proposed action would not be

significant, and therefore, preparation of

an Environmental Impact Statement is

not warranted.

The EA and other documents related

to this proposed action are available for

public inspection and copying at the

NRC Public Document Room, in the

Gelman Building, 2120 L Street N.W.,

Washington, DC 20555.

Notice of Opportunity for Hearing

The Commission hereby provides

notice that this is a proceeding on an

application for a licensing action falling

within the scope of Subpart L, ‘‘Informal

Hearing Procedures for Adjudications in

Materials and Operators Licensing

Proceedings,’’ of the Commission’s

Rules of Practice for Domestic Licensing

Proceedings and Issuance of Orders in

10 CFR Part 2 (54 FR 8269). Pursuant to

§ 2.1205(a), any person whose interest

may be affected by this proceeding may

file a request for a hearing. In

accordance with § 2.1205(c), a request

for a hearing must be filed within thirty

(30) days from the date of publication of

this Federal Register notice. The request

for a hearing must be filed with the

Office of the Secretary either:

(1) By delivery to the Rulemakings

and Adjudications Staff of the Office of

the Secretary at One White Flint North,

11555 Rockville Pike, Rockville, MD

20852; or

(2) By mail or telegram addressed to

the Secretary, U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

Attention: Rulemakings and

Adjudications Staff.

Each request for a hearing must also

be served, by delivering it personally or

by mail to:

(1) The applicant, Umetco Mineral

Corporation, P.O. 1029, Grand Junction,

CO 81502;

(2) The NRC staff, by delivery to the

Executive Director of Operations, One

White Flint North, 11555 Rockville

Pike, Rockville, MD 20852, or

(3) By mail addressed to the Executive

Director for Operations, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555.

In addition to meeting other

applicable requirements of 10 CFR Part

2 of the Commission’s regulations, a

request for a hearing filed by a person

other than an applicant must describe in

detail:

(1) The interest of the requestor in the

proceeding;

(2) How that interest may be affected

by the results of the proceeding,

including the reasons why the requestor

should be permitted a hearing, with

particular reference to the factors set out

in § 2.1205(g);

(3) The requestor’s areas of concern

about the licensing activity that is the

subject matter of the proceeding; and

(4) The circumstances establishing

that the request for a hearing is timely

in accordance with § 2.1205(c).

Any hearing that is requested and

granted will be held in accordance with

the Commission’s ‘‘Informal Hearing

Procedures for Adjudications in

Materials and Operator Licensing

Proceedings’’ in 10 CFR Part 2, Subpart

L.

Dated at Rockville, Maryland, this 30th day

of April 1998.

For the Nuclear Regulatory Commission.

Joseph J. Holonich,

Chief, Uranium Recovery Branch, Division

of Waste Management, Office of Nuclear

Material Safety and Safeguards.

[FR Doc. 98–11980 Filed 5–5–98; 8:45 am]

BILLING CODE 7590–01–P

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.

Pub. L. 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 April 10

through April 24, 1998. The last

biweekly notice was published on April

22, 1998 (63 FR 19964).

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.
25102 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

Administration 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 June 5, 1998, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25103

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendment request:

November 1, 1996, as supplemented by

letters dated October 13, 1997, February

26, 1998, and March 13, 1998.

Description of amendment request:

Associated with a Carolina Power &

Light Company (the licensee)

application to convert from the Current

Technical Specifications (CTS) for the

Brunswick Steam Electric Plant, Units 1

and 2, to Improved Technical

Specifications (ITS), as contained in

Revision 1 of NUREG–1433, ‘‘Standard

Technical Specification General Electric

Plants, BWR/4,’’ the licensee proposed

removing a restriction on a surveillance

test described below.

CTS 4.8.1.1.1.b requires that the

offsite electrical power circuits be

demonstrated OPERABLE, at least once

per 18 months during shut down, by

manually transferring the unit power

supply from the normal circuit to the

alternate circuit. As proposed, ITS SR

3.8.1.8.b will not contain the restriction

to perform the Surveillance ‘‘during

shutdown.’’ Currently, this test is

performed by momentarily paralleling

the 230 kV offsite alternating current

(AC) power sources. The licensee has

stated that paralleling offsite AC power

sources is a controlled evolution and the

increased risk associated with the

performance of this test while the unit

is at power is not significant for the

following reasons: (1) the frequency and

voltages are verified to be within the

required range prior to paralleling the

two offsite AC power sources; (2)

breaker interlocks ensure that the

alternate circuit is connected to the load

prior to opening the preferred circuit;

(3) the test does not result in deenergization of any 4.16 kV emergency

bus and the potential for electrical

perturbations on the grid system is the

same whether performing the transfer

while the unit is at power or while

shutdown; and (4) operating history

indicates that transferring offsite AC

power sources while the units were in

Operational Conditions 1 (power

operation) or 2 (startup) has been

performed satisfactorily without

electrical distribution system

perturbations. The licensee has further

pointed out that Generic Letter 91–04,

‘‘Changes in Technical Specifications to

Accommodate a 24-Month Fuel Cycle,’’

states that licensees may omit the

Technical Specification qualification

that a refueling interval surveillance is

to be performed ‘‘during shutdown.’’

Therefore, consistent with the guidance

provided in Generic Letter 91–04, the

licensee proposed deletion of the

requirement to perform this

Surveillance ‘‘during shutdown’’ as part

of the conversion from CTS 4.8.1.1.1.b

to ITS SR 3.8.1.8.b.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

This change would remove a specific

restriction to perform the verification of the

manual transfer of the unit power supply

from the normal circuit to the alternate

circuit ‘‘during shutdown.’’ The transfer of

the unit power supply from the normal

circuit to the alternate circuit is not an

initiator of any previously analyzed accident.

Therefore, this change does not significantly

increase the frequency of such accidents.

Currently, this test is performed by

momentarily paralleling the 230 kV offsite

AC power sources. Paralleling offsite AC

power sources is a controlled evolution and

the increased risk associated with the

performance of this test while the unit is at

power is not significant for the following

reasons: (1) The frequency and voltages are

verified to be within the required range prior

to paralleling the two offsite AC power

sources; (2) breaker interlocks ensure that the

alternate circuit is connected to the load

prior to opening the preferred circuit; (3) the

test does not result in de-energization of any

4.16 kV emergency bus and the potential for

electrical perturbations on the grid system is

the same whether performing the transfer

while the unit is at power or while

shutdown; and (4) operating history indicates

that transferring offsite AC power sources

while the units were in MODE (Operational

Condition) 1 or 2 has been performed

satisfactorily without electrical distribution

system perturbations. The appropriate plant

conditions for performance of the

Surveillance will continue to be controlled to

assure the potential consequences are not

significantly increased. This control method

has been previously determined to be

acceptable as indicated in Generic Letter 91–

04. Therefore, this change does not

significantly increase the consequences of

any previously analyzed accident.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

This change removes a specific restriction

on the plant conditions for performing a

Surveillance, but does not change the method

of performance. The appropriate plant

conditions for performance of the

Surveillance will continue to be controlled to

assure the possibility for a new or different

kind of accident are not created. This control

method has been previously determined to be

acceptable as indicated in Generic Letter 91–

04. Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously analyzed

accident.

3. Does this change involve a significant

reduction in a margin of safety?

The margin of safety considered in

determining the appropriate plant conditions

for performing the Surveillance will continue

to be controlled to assure that there is no

significant reduction. This control method

has been previously determined to be

acceptable as indicated in Generic Letter 91–

04. Therefore, the change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

University of North Carolina at

Wilmington, William Madison

Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Pao-Tsin Kuo

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendment request: April 3,

1998.

Description of amendment request:

The Carolina Power & Light Company,

licensee for the Brunswick Steam

Electric Plant (BSEP), Unit Nos. 1 and

2, proposed amendments to the

Technical Specifications (TS) to change

the specified total volume of the

condensate storage tank (CST) from

150,000 gallons to 228,200 gallons.

During a recent review of industry

operating experience, the licensee

determined that information contained

in TS 3.5.3.1, Core Spray System (CSS),

and the associated bases regarding water

inventory in the CST was incorrect.

Specifically, the minimum CST volume

requirement contained in TS 3.5.3.1

would not assure the availability of

50,000 gallons of water for the CSS, as

indicated in TS Bases section 3/4.5.3.1

for the CSS.

The licensee has concluded that the

proposed license amendments do not

involve a Significant Hazards

Consideration. In support of this

determination, an evaluation of each of

the three standards set forth in 10 CFR

50.92 is provided below.
25104 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

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 license amendments do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed TS change revises the

minimum CST [Condensate Storage Tank]

water volume required for OPERABILITY of

the Core Spray system (CSS) in

OPERATIONAL CONDITIONS 4 AND 5

when the suppression pool is inoperable. The

proposed change does not alter the operation

of any plant system or component; does not

involve a physical modification to any

structure, system, or component; and does

not affect an initiator to any accident

previously evaluated. The minimum CST

water level is being increased to assure the

availability of 50,000 gallons of water for use

by the CSS. Therefore, the proposed license

amendments do not involve an increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed license amendments will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. This proposed TS

change revises the minimum CST water

volume required for OPERABILITY of the

CSS in OPERATIONAL CONDITIONS 4 and

5 when the suppression pool is inoperable.

The proposed change does not alter the

operation of any plant system or component;

does not involve a physical modification to

any structure, system, or component; and

does not affect an initiator to any accident

previously evaluated. The proposed change

does not add or modify equipment or

components related to the CSS and will,

therefore, not create new failure modes or

common failure modes. The minimum CST

water level is being increased to assure the

availability of 50,000 gallons of water for use

by the CSS. Therefore, the proposed license

amendments do not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety. The proposed license

amendments increase the minimum CST

water level to assure the availability of

50,000 gallons of water for use by the CSS.

These volumes ensure the validity of existing

analyses, and ensure that the existing TS

Bases are satisfied. The proposed change

does not involve a physical modification to

any structure, system, or component, and

does not modify the operation of any existing

equipment. Therefore, the proposed license

amendments do not involve a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendments request involves no

significant hazards consideration.

Local Public Document Room location:

University of North Carolina at

Wilmington, William Madison

Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Pao-Tsin Kuo

(Acting)

Commonwealth Edison Company,

Docket Nos. 50–254 and 50–265, Quad

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of amendment request: March

31, 1998.

Description of amendment request:

Unreviewed Safety Question involving

use of Station Blackout (SBO) diesel

generators (DGs) and use of a mobile

safe shutdown (SSD) battery cart in the

10 CFR part 50, appendix R, Safe

Shutdown Safety Analysis.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The licensee has provided a separate

no significant hazards consideration

determination for the SBO DGs and the

battery cart under this amendment

request. The following is the

determination for the SBO DGs:

(1) No significant increase in the

probability or consequences of an accident

previously evaluated is involved because of

the following:

Two types of previously evaluated

accidents are relevant to this criterion: (1) A

fire; (2) other accident evaluated in the

UFSAR. For these previously evaluated

accidents, the change would not result in an

increase in either their probabilities of

occurrence or the consequences of their

occurrence, for the following reasons.

The use of the SBO DGs in lieu of the

[Emergency Diesel Generators] EDGs does not

change the probability or consequences of a

fire. The likelihood of a fire is unchanged.

Use of the SBO DGs does not significantly

change the fire loading nor introduce

significant new ignition sources. The

consequences of a fire are unchanged because

use of the SBO DGs continues to support the

station’s ability to achieve and maintain

shutdown in the event of a fire.

Use of the SBO DGs for non-fire purposes

is unchanged by use of the SBO DGs for postfire safe shutdown in the event of a fire in

areas requiring alternate shutdown

capability. Accordingly there is no change in

the probability or consequences of a

previously evaluated accident involving the

SBO DGs. Similarly, there is no change to the

probability or consequences of other

accidents that have been previously

evaluated because they are independent of

this change in use of the SBO DGs.

(2) The possibility of a new or different

kind of accident from any accident

previously evaluated is not created because:

The proposed change does not create the

possibility of a new or different kind of

accident from that previously evaluated for

Quad Station. Although the SBO DGs will be

used for a new function, there is no

significant change in the operation of the

SBOs for a non-fire event. Moreover, the

overall use of the SBO DGs as an AC power

source is not significantly different from the

use of the EDGs. The SBO DGs buses provide

power to the same buses that are powered

from the EDGs. No new modes of operation

are introduced by the proposed changes. The

use of the SBO DGs provides a slightly

different but effective method for achieving

and maintaining post-fire safe shutdown for

areas requiring alternate shutdown

capability. As such, the proposed change

does not create the possibility of a new or

different kind of accident.

(3) No significant reduction in the margin

of safety is involved because:

A change in the fire protection program

does not result in a significant reduction in

the margin of safety if the change does not

result in a significant adverse impact on the

plant’s ability to achieve and maintain safe

shutdown in the event of a fire. The proposed

use of the SBO DGs instead of the EDGs to

achieve and maintain safe shutdown within

72 hours change does not significantly affect

the capability or reliability of the equipment

assumed to operate in the safety analysis.

The demonstrated capability and reliability

of the SBO and EDGs are not significantly

different. Indeed, the SBO DGs represent a

safety improvement due to their physical

separation from the postulated fire areas, and

the operational benefits provided by their

greater capacity. Any narrow reduction in

margin associated with the need to manually

start the SBO DGs is offset by the reduction

in manual actions necessary to reduce

electrical loads powered from the EDGs. The

lack of Class 1E qualification for the SBO

DGs is not significant from a safety

perspective because the demonstrated

reliability of the SBO DGs is comparable to

the reliability of the EDGs. The lack of

seismic qualification and single failure

protection do not constitute a significant

reduction in margin since neither of these

attributes is required by Appendix R.

Accordingly, the Commission has already

determined that these attributes are not part

of the Appendix R acceptance criterion. Any

reduction in margin associated with the

greater fuel consumption rate of the SBO DGs

is partially offset by the increased flexibility

in powering equipment to achieve and

maintain post fire safe shutdown.

Additionally, onsite fuel storage and manual

transfer capabilities provide for at least 72

hours of SBO DG operation. Within 72 hours,

deliveries of diesel fuel from offsite supplies

is expected. Therefore, the use of the SBO

DGs as an onsite AC power source for
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25105

equipment necessary to achieve and maintain

post-fire safe shutdown in areas requiring

alternate capabilities does not involve a

significant reduction in margin.

The licensee has evaluated the use of

the mobile SSD battery cart to provide

the power source for the Automatic

Depressurization System (ADS) valves

under certain scenarios where the

valves are needed to achieve cold

shutdown and determined that it does

not involve a significant hazards

consideration for the reasons discussed

below.

(1) No significant increase in the

probability or consequences of an accident

previously evaluated is involved.

The accident previously evaluated is the

postulated fire requiring alternate shutdown

capability. The probability of a previously

evaluated fire is not increased significantly

because the mobile SSD batteries do not

create significant new ignition sources or any

other fire initiators. The consequences of a

previously evaluated fire are not increased

significantly because the mobile SSD

batteries do not significantly increase the fire

loading in the plant, do not interfere with the

plant’s ability to extinguish a fire, and are

fully capable of fulfilling the designed safety

function.

The associated systems related to this

proposed change are not affected in a way

that could impact the initiation of any

accident sequence for the Quad Cities

Station. No modes of operation are

introduced by the proposed change such that

adverse consequences result.

The probability of an accident involving

the use of the mobile SSD batteries would not

be increased significantly by this proposed

use because the use is not significantly

different from the alternative manual

attachment of a power source to the ADS

valves.

The consequences of an accident involving

the use of the mobile SSD batteries are not

increased because the only significant

consequences would be a delay in achieving

cold shutdown and that would have no

different consequences than would a delay

due to an accident related to the currently

used manual power source.

(2) The possibility of a new or different

kind of accident from any accident

previously evaluated is not created.

The proposed change for the Quad Cities

Station does not create the possibility of a

new or different kind of accident from that

previously evaluated. Because the mobile

SSD batteries simply provide a different form

of manually connecting a source of power to

the ADS valves, the use of the mobile SSD

batteries does not present new or different

kinds of accidents related to such manual

actions. Finally, because no new modes of

operation are introduced by the proposed

change, the change does not create the

possibility of a new or different kind of

accident that could be related to new modes

of operation.

(3) No significant reduction in the margin

of safety is involved.

The analytic framework for determining

the extent to which a proposed change affects

the margin of safety has been discussed

above and, so will not be repeated here. In

this case, a review of the proposed changes

shows that they will not have an adverse

impact on the ability to achieve and maintain

safe shutdown. Several features associated

with the use of the mobile SSD batteries

show, as discussed above, that it provides an

effective method for achieving and

maintaining safe shutdown following a fire.

In particular, use of the mobile SSD batteries

reduces the overall complexity of the cold

shutdown repairs required to supply power

to the ADS valves and is familiar to plant

personnel from their training on its use for

other purposes.

Design calculations regarding capabilities

of the mobile SSD batteries show they will

be capable in fulfilling their intended safety

function for their design basis Appendix R

scenario. Reliability of the mobile SSD

batteries will be maintained by augmented

quality standards. This will entail the

conduct of appropriate maintenance and

surveillance which is designed to ensure that

the mobile batteries will function as

intended. Reliability of this power source is

further enhanced by the circumstance that

there are two mobile SSD batteries, thus

permitting one to act as a backup to the other.

Under these circumstances, the margin of

safety for achieving cold shutdown using the

ADS valves is not reduced significantly, if at

all, by the use of non-safety related mobile

SSD batteries to power the ADS valves.

Although safety-related station batteries had

previously been used in this function, the

method for attaching those batteries was

more prone to human error than the method

which has been developed for the mobile

SSD batteries. Moreover, substantial steps

have been taken to provide a high level of

reliability for the mobile SSD batteries.

Overall, therefore, the ability to achieve and

maintain safe shutdown in the event of a fire

has not been reduced by this change in the

source of power to the ADS valves.

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 are satisfied.

Therefore, the NRC staff proposes to

determine that the requested

amendments involve no significant

hazards consideration.

Local Public Document Room location:

Dixon Public Library, 221 Hennepin

Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I. Miller,

Esquire; Sidley and Austin, One First

National Plaza, Chicago, Illinois

60603

NRC Project Director: Stuart A. Richards

Commonwealth Edison Company,

Docket Nos. 50–295 and 50–304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of amendment request: March

30, 1998.

Description of amendment request:

The proposed amendments would

restore the Zion Custom Technical

Specifications (CTS) that had been

replaced with Improved Technical

Specification by a previous amendment

and would reinstate License Conditions

that were deleted by that previous

amendment. The proposed amendment

would also modify the CTS to allow the

use of Certified Fuel Handlers to satisfy

shift staffing requirements and would

change management titles and

responsibilities to reflect the

permanently shutdown organization.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

With a plant permanently shutdown and

defueled the spectrum of accidents and

events that remain credible is significantly

reduced. As discussed below the proposed

changes do not affect the probability or

consequences of any accidents that do

remain credible.

The restoration of the CTS which were

replaced with the ITS by Amendments 178/

165 cannot increase the probability or

consequences of any event or accident

because the amendment was never

implemented. The CTS have been

maintained as the legally binding Technical

Specifications in effect at Zion Station. The

reinstatement of the five License Conditions

deleted by Amendments 178/165 is an

administrative change in that the

requirements contained in the License

Conditions had been relocated elsewhere and

are now being restored exactly as they were

before the amendment was issued. Since the

actual requirements have not changed there

can be no change in the probability or

consequences of any accident or event.

The changes in management titles and

responsibilities will not increase the

probability or consequences of any accident

or event because these changes are

administrative and will not result in any

decrease in the quality of management

applied to Zion Station. The changes are

commensurate with the significant reduction

in site activities, site staffing, and risk to

public health and safety that occurs when an

operational nuclear power plant transitions

to a permanently shutdown and defueled

plant. Responsible individuals will have the

authority to commit the personnel and

resources necessary to fulfill their obligations

for safe storage and handling of nuclear fuel.

The change of position designations will

have no effect on the frequency of occurrence

of accident or event initiators, or on their

consequences.

The changes to allow use of Certified Fuel

Handlers in lieu of personnel licensed in

accordance with 10 CFR part 55 will not

increase the probability or consequences of

an accident or event because the Certified

Fuel Handler Training and Retraining

program (which will be approved by the
25106 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

NRC) has been developed using a Systems

Approach to Training as defined in 10 CFR

55.4. This approach provides assurance that

the Certified Fuel Handlers have the

knowledge, skills, and abilities that are

commensurate with the tasks to be performed

(i.e., the proper monitoring, handling,

storage, and cooling of nuclear fuel).

Therefore the frequency of occurrence of

accident or event initiators is not increased

and the consequences of the accidents or

events are unaffected.

The changes in shift staffing numbers and

crew composition will not increase the

probability or consequences of an accident or

event. These staffing changes are

commensurate with the quantity, complexity,

and hazard level of the activities required for

storage and handling of nuclear fuel. The

elimination of the Shift Control Room

Engineer does not affect any accident or

event initiator or consequence since the

previous specification would not have

required that the position be manned with

both units shut down. The elimination of the

requirement for a Radiation Protection

Person on shift will have no effect on the

frequency of occurrence of accidents or

events, nor on the consequences of the

accident or event.

The changes in verbiage to eliminate any

implication that units are operational will

not increase the probability or consequences

of an accident or event because they are

largely editorial changes and do not increase

the frequency of occurrence of [or] event

initiators, nor do they increase the

consequences.

Therefore this proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Does the change create the possibility of a

new or different kind of accident from any

accident previously evaluated?

The changes proposed by this amendment

do not involve new structures, systems, or

components, or the use of existing structures,

systems, or components in a new manner.

Consequently no new failure mechanisms are

introduced. The design and operation of

structures, systems, or components is

unaffected by:

The restoration of CTS,

The reinstatement of the five License

Conditions deleted by Amendments 178/165,

The changes in management titles and

responsibilities,

The changes to allow use of Certified Fuel

Handlers in lieu of 10 CFR [Part] 55 licensed

personnel,

The changes in shift staffing numbers and

crew composition, or

The changes in verbiage to eliminate any

implication that units are operational.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Does the change involve a significant

reduction in a margin of safety?

One of the License Conditions that would

be reinstated by this amendment establishes

limits that help ensure that the assumptions

of the fuel handling accident analysis remain

valid. License Condition 2.C.(7).b limits the

weight of loads carried over fuel stored in the

spent fuel pool to the weight of a single fuel

assembly plus the tool for moving that

assembly. This weight limit ensures that the

number of fuel rods broken in a fuel handling

accident does not exceed the maximum

number of fuel rods assumed to break in the

accident analysis. Consequently, this change

continues to provide assurance that the

margin of safety involving the number of fuel

rods broken in the accident will not be

reduced.

Therefore, these changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room location:

Waukegan Public Library, 128 N.

County Street, Waukegan, Illinois

60085

Attorney for licensee: Michael I. Miller,

Esquire; Sidley and Austin, One First

National Plaza, Chicago, Illinois

60603

NRC Project Director: Stuart A. Richards

Duke Energy Corporation (DEC), et al.,

Docket Nos. 50–413 and 50–414,

Catawba Nuclear Station, Units 1 and 2,

York County, South Carolina

Date of amendment request: May 27,

1997, as supplemented by a letter dated

April 20, 1998.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications (TS)

of each unit to conform with NUREG–

1431, Revision 1, ‘‘Standard Technical

Specifications—Westinghouse Plants.’’

The Commission had previously issued

a Notice of Consideration of Issuance of

Amendments in the Federal Register on

July 14, 1997 (62 FR 37628) covering all

the proposed changes that were indeed

within the scope of NUREG–1431. In

DEC’s May 27, 1997, submittal, there are

proposed changes that are beyond the

scope of NUREG–1431, which were thus

not covered by the staff’s July 14, 1997,

notice. The following descriptions and

no significant hazard analyses cover

only those beyond-scope changes.

Associated with each change are

administrative/editorial changes such

that the new or revised requirements

would fit into the format of NUREG–

1431.

1. This proposed change affects the

surveillance requirement currently

contained in Sections 4.6.6.1 and

4.6.6.2, regarding the containment valve

injection water system. The requirement

to assure adequate capacity to maintain

system pressure for at least 30 days

would be deleted, the required system

pressure of 16.2 pounds per square inch

gauge (psig) would be replaced with a

surge tank pressure of 36.4 psig, and the

system would be tested at lower

pressures and more restrictive leak

rates.

2. Section 3.9.2.1, regarding the boron

dilution mitigating system, currently

requires both trains to be operable in

Mode 6 (refueling). DEC proposed to

add a note stating that the system may

be blocked during core reloading until

two assemblies are loaded into the core.

Adequate shutdown margin will

continue to be controlled and verified

by other specifications. This blocking

would prevent inadvertent actuation of

the system, which could distract the

operating personnel, but would not

diminish the monitoring function of the

system.

3. DEC proposed to change the

definition of ‘dose equivalent iodine131.’ Subsequently, this proposed

change was withdrawn by letter dated

April 20, 1998.

4. DEC proposed to change Section

3.3.3.6 regarding accident monitoring

instrumentation. Specifically, the

change would (a) increase the time

allowed to return the required number

of channels to operable; and (b) permit

continued operation if one channel is

inoperable given certain conditions are

met, instead of requiring shutdown.

5. DEC proposed to change Section

4.6.4.1 regarding surveillance

requirements for the hydrogen monitors

(combustible gas control). Specifically,

this would eliminate the channel

operational test, and extend the channel

check frequency from once per 12 hours

to once per 31 days.

6. DEC proposed to change Section

3.4.6.1 regarding reactor coolant leakage

detection systems; a system comprising

diverse instruments such as gaseous

radioactivity monitoring, containment

floor and equipment sump monitoring,

etc. In addition to the instruments

specified by this section, the plant has

other installed instruments such as

monitors for humidity, temperature,

etc., which can provide indication for

reactor coolant leakage. Currently, this

specification allows operation up to 30

days if the containment floor and

equipment sump monitoring system is

inoperable. The change would impose a

requirement to perform a precision

water balance of the reactor coolant

system every 24 hours during this

period. The change would also reduce

the number of monitors required

operable provided compensatory

measures are performed or diverse

instruments continue to be available.
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25107

7. DEC proposed to change Section

4.5.4.b, which currently requires

verification of the refueling water

storage tank temperature to be within

the allowed range once per 24 hours if

the outside air temperature is less than

70 degrees or greater than 100 degrees

Fahrenheit. The proposed change would

simply require that the tank temperature

be verified within range every 24 hours

regardless of outside air temperature.

8. DEC proposed to revise Table 3.7–

1, which imposes limits on the

maximum allowable power range

neutron flux high setpoint for various

numbers of inoperable safety valves on

any operating steam generator. The

revision would reduce the setpoints,

making them more conservative.

9. Section 3.7.6, regarding the

condensate storage system, currently

only exists in the Unit 2 TS. DEC

proposed to impose these requirements

also on Unit 1.

10. Several electrical busses and

inverters currently covered by Section

3.8.3.1 are qualified by a footnote,

which specifies the conditions under

which the inverter may be disconnected

from its direct current source. DEC

proposed to delete this footnote because

it is not needed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analyses of the

issue of no significant hazards

consideration for each of the above

proposed changes. The NRC staff has

reviewed the licensee’s analyses against

the standards of 10 CFR 50.92(c). The

NRC staff’s analysis is presented below.

1. Will the change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

For changes 1, 2, 4, 5, 6, 7, 8, 9, and

10, the answer is ‘‘no.’’ The proposed

changes will not affect the safety

function of the subject systems. There

will be no direct effect on the design or

operation of any plant structures,

systems, or components. No previously

analyzed accidents were initiated by the

functions of these systems, and the

systems were not factors in the

consequences of previously analyzed

accidents. Therefore, the proposed

changes will have no impact on the

consequences or probabilities of any

previously evaluated accidents.

2. Will the change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

For changes 1, 2, 4, 5, 6, 7, 8, 9, and

10, the answer is ‘‘no.’’ The proposed

changes would not lead to any hardware

or operating procedure change. Hence,

no new equipment failure modes or

accidents from those previously

evaluated will be created.

3. Will the change involve a

significant reduction in a margin of

safety?

For changes 1, 2, 4, 5, 6, 7, 8, 9, and

10, the answer is ‘‘no.’’ Margin of safety

is associated with confidence in the

design and operation of the plant. The

proposed changes to the TS do not

involve any change to plant design,

operation, or analysis. Thus, the margin

of safety previously analyzed and

evaluated is maintained.

Based on this analysis, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied for each of the proposed

changes. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

York County Library, 138 East Black

Street, Rock Hill, South Carolina

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), et al.,

Docket Nos. 50–413 and 50–414,

Catawba Nuclear Station, Units 1 and 2,

York County, South Carolina

Date of amendment request: April 8,

1998.

Description of amendment request:

The proposed amendments would

revise Section 3.6.5.1 and 4.6.5.1 of the

Technical Specifications (TS) of each

unit to relax ice condenser stored ice

weight requirements by approximately 6

percent. The proposed change is based

mainly on DEC’s gathered data showing

lower sublimation rate than originally

anticipated.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analyses of the

issue of no significant hazards

consideration for the proposed changes.

The NRC staff has reviewed the

licensee’s analyses against the standards

of 10 CFR 50.92(c). The NRC staff’s

analysis is presented below.

1. Will the changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

No. The proposed changes will not

affect the safety function of the ice

condenser in that there will be no

changes to the design or operation of

any plant structures, systems, or

components. No previously analyzed

accidents were initiated by the

functions of the ice condenser, and the

ice condenser will remain fully capable

of performing its design accident

mitigation function. Therefore, the

proposed changes will have no impact

on the consequences or probabilities of

any previously evaluated accidents.

2. Will the changes create the

possibility of a new or difference kind

of accident from any accident

previously evaluated?

No. The proposed changes would not

lead to any hardware or operating

procedure change. Reducing the

required ice weight will not have any

impact on other plant systems that were

assumed to be accident initiators.

Hence, no new equipment failure modes

or accidents from those previously

evaluated will be created.

3. Will the changes involve a

significant reduction in a margin of

safety? No. Margin of safety is

associated with confidence in the design

and operation of the plant; specifically,

the ability of the fission product barriers

to perform their design functions during

and following an accident. The

proposed changes regarding required ice

weight do not involve any change to

plant design, operation, or analysis.

Thus, the margin of safety previously

analyzed and evaluated is maintained.

Based on this analysis, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied for the proposed changes.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room location:

York County Library, 138 East Black

Street, Rock Hill, South Carolina

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), Docket

Nos. 50–369 and 50–370, McGuire

Nuclear Station, Units 1 and 2,

Mecklenburg County, North Carolina

Date of amendment request: May 27,

1997.

Description of amendment request:

The proposed changes would lower the

minimum required diesel generator (DG)

air start receiver pressure from 220 per

square inch gauge (psig) to 210 psig

with a monthly verification, and would

include an allowed outage time of 48

hours for a degraded air receiver

provided the redundant air receiver is

maintained at equal to or greater than

210 psig. These proposed changes are

associated with DEC’s application to

convert to the Improved Technical
25108 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

Specifications. Also, they are

considered less restrictive requirements

because of the lower required minimum

pressure and the allowance of continued

operation with a degraded starting air

system.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration for each change, which is

presented below:

1. (Do the changes) involve a significant

increase in the probability or consequence of

an accident previously evaluated?

The proposed changes provide Actions for

degraded capabilities of the diesel starting air

subsystems for the DG. The proposed Actions

establish limits for the DG starting air

subsystems of 210 psig, (are) allowed to

decrease below the required value for 48

hours(, and are verified every 31 days.) The

Completion Times are based on the amount

of capability remaining, and the time needed

to correct any deficient condition. If the

Completion Times are exceeded, the

specification requires the associated DG to be

declared inoperable immediately, consistent

with the current TS (technical

specifications). Since the new Actions

continue to assure that the associated DG

remains capable of performing its design

safety function, the proposed (changes do)

not significantly affect the probability or

consequences of an accident previously

evaluated.

2. (Do the changes) create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed (changes do) not permit

operation in a new or different mode, or

permit the installation of a new or different

type of equipment. The proposed changes

provide Actions for degraded capabilities of

the DG starting air subsystems. The proposed

Actions establish Conditions, Required

Actions, and Completion Times to be entered

when in a degraded condition. The DG

remains capable of performing its design

safety function. Therefore, the proposed

(changes do) not create the possibility of a

new or different kind of accident from those

previously evaluated.

3. (Do these changes) involve a significant

reduction in a margin of safety?

The proposed (changes do) not

significantly increase the probability or

consequences of an accident previously

evaluated. The changes provide assurance

that timely action will be initiated to restore

DG starting air subsystem when

inoperabilities exist, without unnecessarily

forcing plant shutdown. Based on the limit

for the starting air subsystem for the DG, the

limited time allowed is acceptable to restore

the parameter to within the requirements

without unnecessary plant shutdown.

Therefore, (these 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:

J. Murrey Atkins Library, University

of North Carolina at Charlotte, 9201

University City Boulevard, Charlotte,

North Carolina

Attorney for licensee: Mr. Albert Carr,

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), Docket

Nos. 50–369 and 50–370, McGuire

Nuclear Station, Units 1 and 2,

Mecklenburg County, North Carolina

Date of amendment request: May 27,

1997.

Description of amendment request:

The two proposed changes are

associated with DEC’s application to

convert to the Improved Technical

Specifications and are considered as

administrative changes. The first change

would delete a current requirement to

only verify the refueling water storage

tank temperature once every 24 hours if

the outside air temperature is less than

70 degrees or greater than 100 degrees

Fahrenheit, and would require that the

tank temperature be verified within

range every 24 hours regardless of the

outside air temperature value. The

second change would delete the current

requirement that 32 of 33 hydrogen

igniters be operable on each train, and

would require that 34 igniters per train

to be operable. The actual design

contains 35 igniters per train. This

change would correct an inadvertent

error in the current Technical

Specifications (TS). The number of

igniters was increased to 35 after the

first refueling outage of each unit. This

change would correct the TS to reflect

the requirements stated in Safety

Evaluation Report Supplement 7.

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 for each of the above

proposed changes. The NRC staff has

reviewed the licensee’s analyses against

the standards of 10 CFR 50.92(c). The

NRC staff’s analysis is presented below:

1. Will the change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes will not affect

the safety function of the subject

systems. There will be no direct effect

on the design or operation of any plant

structures, systems, or components. No

previously analyzed accidents were

initiated by the functions of these

systems, and the systems were not

factors in the consequences of

previously analyzed accidents.

Therefore, the proposed changes will

have no impact on the consequences or

probabilities of any previously

evaluated accidents.

2. Will the change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes would not lead

to any hardware or operating procedure

change. Hence, no new equipment

failure modes or accidents from those

previously evaluated will be created.

3. Will the change involve a

significant reduction in a margin of

safety?

Margin of safety is associated with

confidence in the design and operation

of the plant. The proposed changes to

the TS do not involve any change to

plant design, operation, or analysis.

Thus, the margin of safety previously

analyzed and evaluated is maintained.

Based on this analysis, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied for each of the proposed

changes. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

J. Murrey Atkins Library, University

of North Carolina at Charlotte, 9201

University City Boulevard, North

Carolina

Attorney for licensee: Mr. Albert Carr,

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: May 27,

1997.

Description of amendment request:

The proposed change would allow two

charging pumps or safety injection

pumps capable of injecting into the

Reactor Coolant System (RCS) when the

RCS is depressurized and an RCS vent

of at least 4.5 square inches is

established. This proposed change is

associated with the licensee’s

application to convert to the Improved

Technical Specifications and results in

a requirement less restrictive than the

current requirement.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25109

issue of no significant hazards

consideration for each change, which is

presented below:

1. Does the change involve a significant

increase in the probability or consequence of

an accident previously evaluated?

The proposed change will provide an

additional alternative for low temperature

(overpressure) relief capacity when two

charging pumps or safety injection pumps are

capable of injecting into the RCS. The low

temperature (overpressure) protection is not

considered to be an initiator of any analyzed

event, therefore, the proposed change does

not increase the probability of a previously

analyzed event.

The proposed change provides an

equivalent vent size to the existing two open

PORVs (power-operated relief valves).

Therefore, this change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in the manner in which the plant

is operated. The proposed change adds an

additional alternative to overpressure

protection equivalent to the current

requirements. Therefore, the proposed

change will not create the possibility of a

new or different kind of accident than any

previously evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

As described above, the proposed change

adds an additional alternative to

overpressure protection equivalent to the

current requirements. The inclusion of

additional alternatives provides the operating

staff with additional flexibility in meeting

low temperature overpressure protection

requirements. Therefore, the 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:

J. Murrey Atkins Library, University

of North Carolina at Charlotte, 9201

University City Boulevard, Charlotte,

North Carolina

Attorney for licensee: Mr. Albert Carr,

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: March

25, 1998

Description of amendment request:

Revise Technical Specification (TS)

3.9.8.1, ‘‘Shutdown Coolant and Coolant

Circulation High Water Level,’’ and TS

3.9.8.2, ‘‘Shutdown Cooling and Coolant

Circulation Low Water Level,’’ to

change the minimum water level above

the fuel assemblies seated in the reactor

vessel at which the Shutdown Cooling

(SDC) System is required to be

maintained operable, or be in operation.

In addition, TS 3.8.1.2, ‘‘Electric Power

Systems, A.C. Sources, Shutdown,’’ and

Technical Specification Bases 3/4.9.8,

‘‘Shutdown Cooling and Coolant

Circulation,’’ have been changed to

make the wording consistent with TS

3.9.8.1 and TS 3.9.8.2.

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 consequence of any accident?

Response: No.

The operation of the facility in accordance

with this change does not involve an increase

in the probability of any accident.

Changing the water level at which the

Shutdown Cooling (SDC) System is required

to be maintained operable or be in operation

will not increase the probability or

consequences of an accident. The design,

operation, or configuration of the SDC system

will not be changed.

At least one shutdown cooling train will be

in operation to ensure sufficient cooling

capacity is available to remove decay heat

and maintain the water in the reactor

pressure vessel below 140 degree F as

required during the refueling mode.

At least one shutdown cooling train will be

in operation to ensure sufficient coolant

circulation is maintained through the reactor

core to minimize the effects of a boron

dilution incident and prevent boron

stratification. Technical Specification

3.9.10.1, ‘‘Refueling Operations Water

Level—Reactor Vessel Fuel Assemblies,’’ will

be complied with, and therefore, the

assumptions related to iodine removal and

the fuel handling accident will be preserved.

Sufficient time, approximately 1.00 hours,

will be available to the operators to initiate

compensatory measures to preclude the

initiation of core boiling in the unlikely event

SDC should be loss [lost].

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?

Response: No.

The operation of the facility in accordance

with this proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change will not affect the

design, configuration, or operation of the

SDC system, and therefore there are no new

modes of failure introduced.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: No.

Operation of the facility in accordance

with this proposed change will not involve

a significant reduction in a margin of safety.

The calculation of the time to the initiation

of boiling based on 23 feet above the top of

the fuel seated in the reactor vessel, at four

days after shutdown, demonstrates there is

significant time available, approximately 1.00

hour, to the operators within which to take

compensatory measures to preclude the

initiation of boiling. The calculation shows

that based on 23 feet of water above the

reactor flange there is 2.04 hours to the

initiation of boiling. Although there is a

reduction in the time to the initiation of

boiling, compensatory measures could be

taken within a few minutes to restore SDC,

and thus, there is still a significant margin

available to the operators within which to

preclude the initiation of boiling. Thus, the

margin of safety is not significantly reduced.

The time to core uncovery was determined

to be 27.74 hours based on four days after

shutdown and water level twenty-three (23)

feet above the fuel assemblies seated in the

reactor vessel.

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 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 New Orleans Library,

Louisiana Collection, Lakefront, New

Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn, 1400 L Street

N.W., Washington DC 20005–3502

NRC Project Director: John N. Hannon

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

20, 1998.

Description of amendment request:

The proposed amendment requests

editorial changes to the Improved

Technical Specifications (ITS) Safety

Limits and Administrative Controls to

replace the titles of the Senior Vice

President, Nuclear Operations (SVPNO)

and the Vice President, Nuclear

Production (VPNP) with the position of

Chief Nuclear Officer (CNO). The CNO

combines the duties of the SVPNO and

VPNP as currently described in ITS and

is required to be an officer of the

company. The proposed change is
25110 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

intended to allow upgrading the

position of the corporate officer

responsible for overall nuclear

operations without limiting the title.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

Does not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed amendment does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because the deletion and updating

of individual titles does not affect plant

operation. No design basis accidents are

affected by the proposed administrative and

editorial changes and, as such, there are no

physical changes to the facility or its

operation.

Does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed ITS changes are

administrative and editorial in nature. No

changes to the facility structures, systems

and components or their operation will

result. The design and design basis of the

facility remain unchanged. The plant safety

analyses remain current and accurate. No

new or different failure mechanisms are

introduced. Therefore, the possibility of a

new or different kind of accident from any

accident previously evaluated is not

introduced.

Does not involve a significant reduction in

the margin of safety.

The proposed ITS changes are

administrative and editorial in nature. The

proposed safety margins established through

the design and facility license including the

Improved Technical Specifications remain

unchanged. In addition, the proposed

amendment ensures continued emphasis and

assignment of responsibility for overall

nuclear safety. Therefore, all margins of

safety are maintained.

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

34428

Attorney for licensee: R. Alexander

Glenn, General 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

20, 1998.

Description of amendment request:

The proposed amendment would

change the Inservice Inspection Program

described in Improved Technical

Specification (ITS) 5.6.2.8.c. This ITS

currently states that the reactor coolant

pump (RCP) motor flywheels will be

inspected during the ‘‘Spring 1998

refueling outage,’’ which would have

been refueling outage 11. Due to a recent

17-month extended outage, refueling

outage 11 has been deferred until Fall

1999. The proposed change is intended

to accurately reflect the new refueling

outage 11 schedule.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

The proposed change will not significantly

increase the probability or consequences of

an accident previously evaluated.

The safety function of the RCP flywheels

is to provide a coastdown period during

which the RCPs would continue to provide

reactor coolant flow to the reactor after loss

of power to the RCPs. The maximum loading

on the RCP motor flywheel results from

overspeed following a large loss of coolant

accident (LOCA). The estimated maximum

obtainable speed in the event of a Reactor

Coolant System piping break was established

conservatively. The proposed one-time

editorial change to remove the words ‘‘Spring

1998 refueling outage’’ and replace them

with ‘‘to coincide with Refueling Outage

11R’’ does not affect that analysis. The

proposed change in dates is editorial in that

it merely reflects the new date for cycle 11.

The usage time for the flywheels is bounded

by the original estimates. The proposed

editorial change does not affect the amount

of radioactive material available for release or

modify any systems used for mitigation of

such releases during accident conditions.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

The proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed editorial change will not

change the design, configuration, or method

of operation of the plant. Therefore, the

proposed change will not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed change will not involve a

significant reduction to any margin of safety.

The proposed Amendment is an editorial

change to reflect that CR–3’s operating cycle

is not ending in spring 1998, but in fall 1999.

The proposed change does not affect the

methods of inspection or its acceptance

criteria. Therefore, the margins of safety

defined in RG [Regulatory Guide] 1.14 are not

changed.

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

34428

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida

Power Corporation, MAC–A5A, P.O.

Box 14042, St. Petersburg, Florida

33733–4042

NRC Project Director: Frederick J.

Hebdon

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: April 15,

1998.

Description of amendment request:

The proposed amendment would

update the existing pressuretemperature curves with new curves

with values from 18 to 32 effective full

power years based on the testing and

analysis of reactor pressure vessel

surveillance materials.

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 amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The pressuretemperature limits are not derived from

Design Basis Accident (DBA) analyses. They

are prescribed by the ASME B&PV Code and

10 CFR part 50 appendices G and H as

restrictions on normal operation to avoid

encountering pressure, temperature, and

temperature rate of change conditions that

might cause undetected flaws to propagate

and cause nonductile failure of the reactor

coolant pressure boundary.

(2) The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The amendment will

merely update the pressure-temperature

curves (and associated SRs and Bases)

already existing in the plant Improved

Technical Specifications to provide limits

from 18 to 32 EFPY of operation, which are

based upon evaluation and analysis of actual

in-vessel material specimens, per 10 CFR part
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25111

50, appendices G and H. The pressuretemperature curves are established to the

requirements of 10 CFR part 50, appendix G

to assure that brittle fracture of the reactor

vessel is prevented.

(3) The proposed amendment will not

involve a significant reduction in a margin of

safety. 10 CFR part 50, appendix G specifies

fracture toughness requirements to provide

adequate margins of safety during operation

over the service lifetime. The values of

adjusted reference temperature and upper

shelf energy determined as a result of the 10

CFR part 50, appendices G and H analysis are

expected to remain within the limits of

Regulatory Guide 1.99, Revision 2 and

appendix G of 10 CFR part 50 (less than 200°

F and greater than 50 ft-lbs respectively) for

at least 32 EFPY of operation.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Cedar Rapids Public Library, 500 First

Street, SE., Cedar Rapids, IA 52401

Attorney for licensee: Jack Newman, Al

Gutterman, Morgan, Lewis & Bockius,

1800 M Street, NW., Washington, DC

20036–5869

Acting NRC Project Director: Richard P.

Savio

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: March

27, 1997.

Description of amendment request:

The proposed amendment, included as

part of the proposed conversion from

the current Technical Specifications

(TS) to improved TS, would establish

Allowable Values for the

instrumentation included in Section 3.3,

as a result of the plant-specific

application of the General Electric

Instrument Setpoint Methodology to the

Cooper Nuclear Station (CNS).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change in selected

Allowable Values for the instrumentation

included in proposed Section 3.3 of the

Technical Specifications is the result of

application of the CNS instrumentation

setpoint methodology. This methodology

incorporates the guidance of ISA

Recommended Practice ISA–RP67.04, Part II,

‘‘Methodologies for the Determination of

Setpoints for Nuclear Safety-Related

Instrumentation,’’ September 1994.

Application of this methodology results in

instrumentation selected Allowable Values

which more accurately reflect total

instrumentation loop accuracy as well as that

of test equipment and setpoint drift between

Surveillances. The proposed change will not

result in any hardware changes. The

instrumentation included in proposed

Section 3.3 of the Technical Specifications is

not assumed to be an initiator of any

analyzed event. Existing operating margin

between plant conditions and actual plant

setpoints is not significantly reduced due to

this change. As a result, the proposed change

will not result in unnecessary plant

transients.

The role of the proposed Section 3.3

instrumentation is in mitigating and thereby

limiting the consequences of accidents. The

Allowable Values have been developed to

ensure that the design and safety analysis

limits will be satisfied. The methodology

used for the development of the Allowable

Values ensures the affected instrumentation

remains capable of mitigating design basis

events as described in the safety analyses and

that the results and consequences described

in the safety analyses remain bounding.

Additionally, the proposed change does not

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. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change is the result of

application of the CNS instrumentation

setpoint methodology and do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. This is based on the fact that the

method and manner of plant operation is

unchanged. The use of the proposed

Allowable Values does not impact safe

operation of CNS in that the safety analysis

limits will be maintained. The proposed

Allowable Values involve no system

additions or physical modifications to

systems in the station.

These Allowable Values were developed

using a methodology to ensure the affected

instrumentation remains capable of

mitigating accidents and transients. Plant

equipment will not be operated in a manner

different from previous operation, except that

setpoints may be changed. Since operational

methods remain unchanged and the

operating parameters have been evaluated to

maintain the station within existing design

basis criteria, no different type of failure or

accident is created.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change does not involve a

reduction in a margin of safety. The proposed

changes have been developed using a

methodology to ensure safety analysis limits

are not exceeded. As such, this proposed

change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE

68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499,

Columbus, NE 68602–0499

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: March

27, 1997.

Description of amendment request:

The proposed amendment, included as

part of the proposed conversion from

the current Technical Specifications

(CTS) to the improved Technical

Specifications (ITS), would add an

additional action statement to a limiting

condition for operation (LCO). The LCO

is in the Improved Standard Technical

Specifications (ISTS, NUREG–1433,

Revision 1) 3.6.2.3 on the residual heat

removal suppression pool cooling

subsystems. The requirements in the

proposed ITS 3.6.2.3 on the subsystems

do not exist in the CTS. The Action B

for ITS 3.6.2.3 would require that if the

two such subsystems were inoperable,

one subsystem would have to be

restored to operability within 8 hours or

the plant would be in ITS 3.0.3. ITS

3.0.3 governs plant operation if an LCO

(i.e., ISTS 3.6.2.3) and the associated

action statement are not met (i.e., Action

B to ISTS 3.6.2.3).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change provides more

stringent requirements for operation of the

facility. These more stringent requirements

do not result in operation that will increase

the probability of initiating an analyzed event

and do not alter assumptions relative to (the)

mitigation of an accident or transient event.

The more restrictive requirements continue

to ensure * * * systems, and components

((i.e., the residual heat removal suppression

pool cooling subsystems)) are maintained

consistent with the safety analyses and

licensing basis. Therefore, this (the proposed)
25112 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

change does not involve a significant (an)

increase in the probability or consequences

of any accident previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in the methods governing normal

plant operation. The proposed change does

impose different requirements. However, this

change is consistent with the assumptions in

the safety analyses and licensing basis. Thus,

this change does not create the possibility of

a new or different kind of accident from any

accident previously evaluated is not created.

3. Does this change involve a significant

reduction in a margin of safety?

The imposition of more restrictive

requirements either has no impact on or

increases the margin of plant safety. As

provided in the discussion of the change,

each change in this category (i.e., more

restrictive requirements) is, by definition,

providing additional restrictions to enhance

plant safety. The change maintains

requirements (systems and components)

within the safety analyses and licensing

basis. Therefore, this change does not involve

a significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE

68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499,

Columbus, NE 68602–0499

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: March

27, 1997.

Description of amendment request:

The proposed amendment, included as

part of the proposed conversion from

the current Technical Specifications

(CTS) to the improved Technical

Specifications (ITS), would add an

additional test (i.e., water and sediment

content within limits) of diesel fuel oil

that could be used in place of a current

test (i.e., clear and bright appearance

with proper color) in the diesel fuel oil

testing program. The current tests are

listed in CTS 4.9.A.2.d/e. The testing

program will be in the new ITS 5.5.9.

The additional test is change number 25

to Section 5.0 of the Improved Standard

Technical Specifications (NUREG–1433,

Revision 1).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change provides more

stringent requirements for operation of the

facility. (This) more stringent (requirement)

do(es) not result in operation that will

increase the probability of initiating an

analyzed event and do(es) not alter

assumptions relative to (the) mitigation of an

accident or transient event. The more

restrictive (requirement) continue(s) to

ensure * * * systems and components (i.e.,

the diesel generators) are maintained

consistent with the safety analyses and

licensing basis. Therefore, the proposed

change does not involve an increase in the

probability or consequences of any accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in the methods governing normal

plant operation. However, this change is

consistent with the assumptions in the safety

analyses and licensing basis. Thus, this

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated is not created.

3. Does this change involve a significant

reduction in a margin of safety?

The imposition of more restrictive

requirements either has no impact on or

increases the margin of plant safety. As

provided in the discussion of the change,

each change in this category (i.e., a more

restrictive requirement) is, by definition,

providing additional restrictions to enhance

plant safety. The change maintains (systems

and components) within the safety analyses

and licensing basis. Therefore, this change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE

68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499,

Columbus, NE 68602–0499

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: March

27, 1997.

Description of amendment request:

The proposed amendment, included as

part of the proposed conversion from

the current Technical Specifications

(TS) to improved TS for the Cooper

Nuclear Station (CNS), would relocate

the Trip Level Settings for the Rod

Block Monitor from Table 3.2.C of the

current TS to the Core Operating Limits

Report. Also, details relating to the

Alternate Shutdown system design and

operation are proposed to be relocated

from current TS 3.2.I and 4.2.I to the

improved TS Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the three criteria of 10 CFR 50.92(c), and

has determined the following:

The proposed changes relocate certain

details from the Technical

Specifications to the Bases and the Core

Operating Limits Report (COLR). The

Bases and the COLR containing the

relocated information will be

maintained in accordance with 10 CFR

50.59. In addition, the Bases and COLR

are subject to the applicable change

control provisions of Chapter 5.0,

Administrative Controls’’, of the

proposed improved Technical

Specifications. Since any changes to the

Bases or the COLR will be evaluated per

the requirements of 10 CFR 50.59 or

other applicable change control

provisions, no increase in the

probability or consequences of an

accident previously evaluated will

result. Therefore, these changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes do not involve

any physical alterations to the plant (no

new or different type of equipment will

be installed), or changes in the methods

governing normal plant operation. The

proposed changes will not impose or

eliminate any requirements, and

adequate control of the information will

be maintained. Thus, these changes do

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes will not reduce

a margin of safety because they have no

impact on any safety analysis

assumptions. In addition, the details to

be transposed from the TS to the Bases
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25113

and the COLR are unchanged. Since any

future changes to these details in the

Bases or the COLR will be evaluated per

the requirements of 10 CFR 50.59 or

other applicable change control

provisions, no reduction in a margin of

safety will result. As such, these

proposed changes do not involve a

significant reduction in a margin of

safety.

Based on the above discussion, 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:

Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE

68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499,

Columbus, NE 68602–0499

NRC Project Director: John N. Hannon

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 8,

1998.

Description of amendment request:

The proposed change would revise

Technical Specifications (TSs) 4.4.5.3,

Steam Generators—Inspection

Frequencies, and 3.4.6.2.c, Reactor

Coolant System (RCS) Leakage, and the

associated bases to accommodate fuel

cycles of up to 24 months with respect

to the allowed time interval between

steam generator inservice inspections.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Extending Surveillance Requirement (SR)

4.4.5.3 to accommodate a 24 month cycle for

inspection of steam generator tubes structural

integrity, as well as, imposing a more

restrictive Limiting Condition for Operation

(TS 3.4.6.2.c) for reactor coolant system

leakage through Category C–2 steam

generators, will neither exacerbate nor

significantly increase the probability or

consequences of an accident previously

evaluated in the Seabrook Station [updated

final safety analysis report] UFSAR.

The proposed changes to SR 4.4.5.3 do not

alter the intent or method by which the

surveillances are conducted, do not involve

physical changes to the plant, do not alter the

way structures, systems or components

(SSCs) function, and do not modify the

manner in which the plant is operated.

The proposed change to TS 3.4.6.2.c

imposes more restrictive limits on plant

operations due to RCS leakage through steam

generators. The proposed change does not

involve physical changes to the plant or alter

the way a SSC functions.

The proposed changes to SR 4.4.5.3 and TS

3.4.6.2.c, and their associated Bases, will not

adversely affect the ability of the steam

generators to perform their intended safety

function. Furthermore, the proposed changes

do not adversely affect the physical

protective boundaries of the plant. The

proposed changes do not affect accident

initiators or precursors and do not alter the

design assumptions, conditions,

configuration of the facility or the manner in

which the plant is operated. The proposed

changes do not alter or prevent the ability of

SSCs to perform their intended function to

mitigate the consequences of an initiating

event within the acceptance limits assumed

in the Updated Final Safety Analysis Report

(UFSAR). The proposed changes are

administrative in nature and do not change

the level of programmatic controls or the

procedural details associated with

aforementioned surveillance requirements.

While the proposed changes will lengthen

the interval between surveillances, the

increase in interval has been evaluated; and

based on the reviews of the steam generator

tube eddy current test (ECT) inspections, it

is concluded that the wear growth rate of the

only active degradation mechanism (AntiVibration Bar (AVB) wear) identified to date

at Seabrook Station is such that sufficient

margin exists between the plugging criteria

and structural limit such that no tubes are

predicted to exceed the structural limit even

with the longer surveillance interval.

Since there are no changes to previous

accident analyses, the radiological

consequences associated with these analyses

remain unchanged, therefore, the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated. Therefore, the

proposed changes will not significantly

increase the probability or consequences of

any previously analyzed accident.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed changes to TS 3.4.6.2 and SR

4.4.5.3, and associated Bases, do not alter the

design assumptions, conditions,

configuration of the facility or the manner in

which the plant is operated. There are no

changes to the source term, containment

isolation or radiological release assumptions

used in evaluating the radiological

consequences in the Seabrook Station

UFSAR. Existing system and component

redundancy is not being changed by the

proposed changes. The proposed changes

have no impact on component or system

interactions. The proposed changes are

administrative in nature and do not change

the level of programmatic controls and

procedural details associated with the

aforementioned surveillance requirements.

Therefore, since there are no changes to the

design assumptions, conditions,

configuration of the facility, or the manner in

which the plant is operated and surveilled,

the proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed change ( ) to the surveillance

intervals for SR 4.4.5.3 is still consistent with

the basis for the interval. The intent or

method of performing the surveillances

remains unchanged. The more restrictive

limit for leakage through any one steam

generator placed in Category C–2, as well as,

the requirement to do an engineering

assessment of steam generator tube integrity,

provides additional margin of ensuring safe

plant operation.

In addition, there is no adverse affect on

equipment design or operation and there are

no changes being made to the Technical

Specification required safety limits or safety

system settings that would adversely affect

plant safety. The proposed changes are

administrative in nature and do not change

the level of programmatic controls and

procedural details associated with the

aforementioned surveillance requirements.

While the proposed changes will lengthen

the interval between surveillances, the

increase in interval has been evaluated; and

based on the reviews of the steam generator

tube ECT inspections, it is concluded that the

wear growth rate of the only active

degradation mechanism (AVB wear)

identified to date at Seabrook Station is such

that sufficient margin exists between the

plugging criteria and structural limit such

that no tubes are predicted to exceed the

structural limit even with the longer

surveillance interval. Therefore, extension of

the current surveillance intervals to

accommodate a 24 month cycle will not

significantly degrade the ability, the

availability or the reliability of the steam

generators to perform their intended safety

function, thus, it is concluded that there is

no significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room location:

Exeter Public Library, Founders Park,

Exeter, NH 03833

Attorney for licensee: Lillian M. Cuoco,

Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

PO Box 270, Hartford, CT 06141–0270

NRC Project Director: Cecil O. Thomas

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: April 6,

1998.

Description of amendment request:

The proposed amendment will modify
25114 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

the Technical Specifications (TSs) by (1)

adding a surveillance requirement to

verify pressurizer heater capacity to TS

3.4.4, ‘‘Reactor Coolant System—

Pressurizer,’’ (2) moving the

identification of the location of the

containment air temperature detectors

from the surveillance requirements

portion of TS 3.6.1.5, ‘‘Containment

Systems—Air Temperature,’’ to the TS

Bases for Containment Systems, Section

3/4.4.6.1.5, ‘‘Air Temperature,’’ and (3)

modifying the action statements and

surveillance requirements of TS 3.7.1.5,

‘‘Plant Systems—Main Steam Isolation

Valves.’’ The TS Bases would also be

updated to include the list of

containment air temperature detectors

and reflect the proposed changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to add a surveillance

requirement (SR) 4.4.4.2 to verify pressurizer

heater capacity will help ensure the

pressurizer will be able to function as

designed to maintain Reactor Coolant System

pressure. There will be no effect on any

design basis accident previously evaluated or

on any equipment important to safety.

Therefore, the proposed change will not

result in a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to modify the

wording of SR 4.6.1.5 and to relocate the list

of containment air temperature detectors

from SR 4.6.1.5 to the Bases will not affect

the Technical Specification limit for

containment temperature or the frequency of

verification of this limit. The proposed

changes do not alter the way any structure,

system, or component functions. The initial

assumption for containment temperature

used in the design basis accident analysis

will remain the same. There will be no affect

on any design basis accident previously

evaluated or on any equipment important to

safety. Therefore, the proposed changes will

not result in a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to the action

statements and surveillance requirements of

Technical Specification 3.7.1.5 will not affect

the operability requirements of the main

(steamline) isolation valves (MSIVs). There

will be no effect on any design basis accident

previously evaluated or on any equipment

important to safety. Therefore, the proposed

changes will not result in a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed changes have no adverse

effect on any of the design basis accidents

previously evaluated or on any equipment

important to safety. Therefore, the License

Amendment Request does not impact the

probability of an accident previously

evaluated nor does it involve a significant

increase in the consequences or an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes will not alter the

plant configuration (no new or different type

of equipment will be installed) or require any

new or unusual operator actions. They do not

alter the way any structure, system, or

component functions and do not alter the

manner in which the plant is operated. The

proposed changes do not introduce any new

failure modes. Therefore, the proposed

changes will 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 will add SR 4.4.4.2

to verify pressurizer heater capacity, relocate

the list of containment temperature detectors

used to verify containment temperature from

SR 4.6.1.5 to the associated Bases, and

modify the action statements and

surveillance requirements of Technical

Specification 3.7.1.5.

These changes will have no adverse effect

on equipment important to safety. This

equipment will continue to function as

assumed in the design basis accident

analysis. Therefore, there will be no

significant reduction in the margin of safety

as defined in the Bases for the technical

Specifications affected by these proposed

changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Learning Resources Center, Three

Rivers Community-Technical College,

574 New London Turnpike, Norwich,

Connecticut, and the Waterford

Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M. Cuoco,

Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut

NRC Deputy Director: Phillip F. McKee

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: April 13,

1998

Description of amendment request:

The proposed amendment would

change the Technical Specifications

(TSs) by adding a new TS 3.5.5,

‘‘Emergency Core Cooling Systems—

Trisodium Phosphate (TSP).’’ Also, the

surveillance requirements in TSs

4.5.2.c.3 and 4.5.2.c.4 would be

relocated to new TS 3.5.5 as TS 4.5.5.1

and TS 4.5.5.2, respectively. The

applicable TS Index page and Bases

sections will be updated to reflect the

proposed changes.

Changes to the current requirements

for the TSP are also proposed. The TSP

requirements in TS 4.5.2.c.3 would

become the limiting conditions for

operation in the new TS; the amount of

TSP required would increase from

‘‘equal to or greater than 110 cubic feet’’

to ‘‘equal to or greater than 282 cubic

feet’’ based on the new calculations; the

applicability would be expanded to

include all of Mode 3; the action

statement would allow 48 hours to

restore the TSP volume; and changes

would also be made to the required tests

and specific details would be relocated

to the applicable TS Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to relocate the

current trisodium phosphate (TSP)

dodecahydrate Technical Specification

requirements from the surveillance

requirements for the Emergency Core Cooling

System to a new TSP Technical Specification

will not change the requirement to store TSP

inside containment. The proposed changes

will require a large quantity of TSP to be

stored inside containment. This large

quantity, based on a recently revised

calculation, will ensure sufficient TSP is

available for containment sump water pH

control. These proposed changes do not alter

the way any structure, system, or component

functions. There will be no adverse effect on

any design basis accident previously

evaluated, on any equipment important to

safety, or o n the radiological consequences

of any design basis accident. Therefore, this

License Amendment Request does not impact

the probability of an accident previously

evaluated nor does it involve a significant

increase in the consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change to increase the TSP

volume stored inside containment will

require two of the wire mesh TSP baskets

inside containment to be replaced by two

new and larger wire mesh baskets. The

design of the new baskets has been evaluated

and it is consistent with the requirements for

equipment installed in containment. The

replacement of the two wire mesh baskets
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25115

will not result in any significant change in

plant configuration and will not require any

new or unusual operator actions. It will alter

the way any structure, system, or component

functions and does not alter the manner in

which the plant is operated. It will not

introduce any new failure modes. Therefore,

the proposed changes will 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 will relocate the

current Technical Specification requirements

for TSP to a new Technical Specification.

The minimum required volume will be

increased to reflect the results of a new

calculation performed to support the current

requirement to raise containment sump pH

[equal to or greater than] 7.0. These changes

will have no adverse effect on equipment

important to safety. This equipment will

continue to function as assumed in the

design basis accident analysis. Therefore,

there will be no significant reduction of the

margin of safety as defined in the Bases for

the Technical Specifications affected by these

proposed changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Learning Resources Center, Three

Rivers Community-Technical College,

574 New London Turnpike, Norwich,

Connecticut, and the Waterford

Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M. Cuoco,

Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut

NRC Deputy Director: Phillip F. McKee

Northern States Power Company,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: April 11,

1997 (supersedes July 26, 1996,

application)

Description of amendment request:

The proposed amendment would

modify the Monticello Technical

Specifications (TS) sections 3.6.C,

Coolant Chemistry, and 3/4.17.B,

Control Room Emergency Filtration

System. The changes were proposed to

establish TS requirements consistent

with modified analysis inputs used for

the evaluation of the radiological

consequences of the main steam line

break accident.

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 amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

A limit is established in the plant

Technical Specifications for steady state

radioiodine concentration in the reactor

coolant to ensure that in the event of a

release of radioactive material to the

environment due to a postulated high energy

line break up to and including a design basis

Main Steam Line Break Accident, radiation

doses are maintained within the guidelines of

10 CFR part 100. The steady state radioiodine

concentration in the reactor coolant is an

input for analysis of the radiological

consequences of an accident due to a Main

Steam Line Break outside of containment and

postulated high energy line breaks. In

addition, requirements are established in the

Technical Specifications for control room

habitability. During an accident, the control

room emergency filtration system provides

filtered air to pressurize the Control Room to

minimize the activity, and therefore the

radiological dose, inside the control room.

A change is proposed for the steady state

radioiodine concentration. This value is

conservative with respect to the value used

in the Main Steam Line Break dose

consequences analysis and is consistent with

the dose consequences evaluation of a

postulated Reactor Water Cleanup (RWCU)

line break. Changes are proposed to the

limiting conditions for operation and

surveillance requirements for the Control

Room Emergency Filtration Train iodine

removal efficiency. These changes are

consistent with the inputs used in the

analysis of the radiological consequences of

the postulated RWCU line break and the

Main Steam Line Break Accident. These

proposed requirements maintain operating

restrictions for analytical inputs used in the

analysis of the Main Steam Line Break

Accident. Evaluation of these events has

demonstrated that the postulated radiological

consequences will remain within the

licensing basis established in the AEC

[Atomic Energy Commission] Provisional

Operating License Safety Evaluation Report,

dated March 18, 1970, thus the proposed

changes do not result in an increase in the

consequences of previously evaluated

accidents.

The analysis of the Main Steam Line Break

Accident performed using a reactor coolant

radioiodine concentration of 2 (microcuries)/

gm dose equivalent Iodine-131 and a control

room ventilation filter efficiency consistent

with the proposed Technical Specifications

changes demonstrated that radiological

consequences of the Main Steam Line Break

are not changed significantly. The

radiological consequences of the Main Steam

Line Break Accident remain within the

exposure guidelines of 10 CFR part 100 and

10 CFR part 50 appendix A, General Design

Criterion 19. The offsite dose consequences

remain bounded by the licensing basis

provided in the AEC Provisional Operating

License Safety Evaluation Report, dated

March 18, 1970. The control room doses

calculated for the hot standby Main Steam

Line Break Accident using the TID–14844

dose conversion factors remain bounded by

the dose consequences of the comparable

design basis loss of coolant accident.

The evaluation of the postulated RWCU

line break, performed using a reactor coolant

radioiodine concentration of 0.25

(microcurie)/gm dose equivalent Iodine-131

and a control room ventilation filter

efficiency consistent with the proposed

Technical Specifications changes,

demonstrated that the radiological

consequences of this event remain within the

exposure guidelines of 10 CFR part 100 and

10 CFR part 50 Appendix A, General Design

Criterion 19. The offsite dose consequences

remain bounded by the Main Steam Line

Break as established in the licensing basis

provided in the AEC Provisional Operating

License Safety Evaluation Report, dated

March 18, 1970.

The proposed Technical Specification

changes do not introduce new equipment

operating modes, nor do the proposed

changes alter existing system interrelationships. The proposed changes do not

introduce new failure modes. The system

improvements to reduce bypass leakage

during postulated accidents do not have an

adverse effect on control room habitability.

Therefore, this amendment will not cause a

significant increase in the probability of an

accident previously evaluated for the

Monticello plant.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed Technical Specification

changes do not introduce new equipment

operating modes, nor do the proposed

changes alter existing system interrelationships. Operator action to mitigate the

consequences of the postulated RWCU line

break is conservative based on the very

limited action required by the operator to

close the containment isolation valves and

the availability of control room indications to

alert the operator to the postulated break. The

use of a ten (10) minute operator response

time to take manual actions in response to

postulated events is consistent with

Monticello’s licensing basis for similar

events. The use of operator actions and all

available equipment is consistent with

current regulatory guidance for mitigating the

consequences of postulated line breaks.

The proposed change to the specification

for reactor coolant dose equivalent

radioiodine is conservative with respect to

the re-evaluation of the Main Steam Line

Break Accident for the more conservative hot

standby initial condition for the postulated

accident. The proposed change to the

specification for reactor coolant dose

equivalent radioiodine is consistent with the

postulated high energy line break of a Reactor

Water Cleanup line. The proposed changes to

the limiting conditions for operation and
25116 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

surveillance requirements for the control

room emergency filtration train iodine

removal efficiency are consistent with the

inputs used in the evaluation of the

radiological consequences of the postulated

RWCU line break and the Main Steam Line

Break Accident. The system improvements to

reduce bypass leakage during postulated

accidents do not have an adverse effect on

control room habitability. Therefore, the

proposed amendment will not create the

possibility of a new or different kind of

accident.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

Surveillance data has demonstrated the

proposed requirements are within the current

capability of the facility. The proposed

changes maintain margins of safety. These

proposed requirements maintain operating

restrictions for analytical inputs used in the

analysis of the bounding postulated high

energy line break of a Reactor Water Cleanup

line and the Main Steam Line Break

Accident. The proposed change to the

specification for reactor coolant dose

equivalent radioiodine is conservative with

respect to the re-evaluation of the Main

Steam Line Break Accident for the more

conservative hot standby initial condition for

the postulated accident. The proposed

change to the specification for reactor coolant

dose equivalent radioiodine is consistent

with the postulated high energy line break of

a Reactor Water Cleanup line. The evaluation

of these postulated events determined that

the radiological consequences remain within

the exposure guidelines of 10 CFR part 100

and of 10 CFR part 50 Appendix A, General

Design Criterion 19. The proposed changes to

the limiting conditions for operation and

surveillance requirements for the control

room emergency filtration train iodine

removal efficiency provide assurance that the

system will perform at the filter efficiency as

used in the evaluation of the radiological

consequences of the postulated events.

Therefore, the proposed amendment will not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Cynthia A.

Carpenter

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment request: April 10,

1998.

Description of amendment request:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant Unit Nos. 1 and 2

to revise TS 6.2.2.g and 6.3 to change

the name of the Operations Manager to

Operations Director and to change the

requirement for the Operations Director

to hold a senior reactor operator (SRO)

license.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change to revise the title of

the Operations Manager to Operations

Director is an administrative change that

clarifies the Technical Specification (TS) to

reflect current position titles.

The proposed change provides assurance

that the Operations Director will continue to

have knowledge of pressurized water reactor

(PWR) operation and emergency event

mitigation. The proposed change does not

detract from the Operations Director’s ability

to perform his primary responsibilities. In

this case, by having previously held a senior

reactor operator (SRO) license, the

Operations Director has achieved the

necessary training, skills, and experience to

fully understand the operation of plant

equipment and the watch requirements for

operators. In summary, the proposed change

does not affect the ability of the Operations

Director to provide the plant oversight

required of his position.

Additionally, another off-shift individual

that holds an SRO license for Diablo Canyon

Power Plant (DCPP) directs the licensed

activities of licensed operators (an

Operations middle manager) will have

specific knowledge of operation and

emergency event mitigation at DCPP. This

will assure that the change in qualification of

the Operations Director does not affect the

probability of an operator initiating an

accident or increasing the consequences of an

accident due to improper direction from

management. The training and qualification

programs for operators on shift will not be

affected by the proposed changes.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change to revise the title of

the Operations Manager to Operations

Director is an administrative change that

clarifies the TS to reflect current position

titles.

The proposed change to TS 6.2.2g. and 6.3

do not affect the design or function of any

plant system, structure, or component, nor

does it change the way plant systems are

operated. It does not affect the performance

of NRC licensed operators since the proposed

changes do not impact the training or

qualification of any operator on shift.

Operation of the plant in conformance with

TS and other license requirements will

continue to be supervised by personnel who

hold an SRO license. The proposed change

to TS 6.2.2g and 6.3 ensures that the

Operations Director will be a knowledgeable

and qualified individual by requiring the

individual to have held an SRO license at a

PWR.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change to revise the title of

the Operations Manager to Operations

Director is an administrative change that

clarifies the TS to reflect current position

titles.

The proposed change involves an

administrative control that is not related to

the margin of safety. The proposed change

does not reduce the level of knowledge or

experience required of an individual who

fills the Operations Director position, nor

does it affect the conservative manner in

which the plant is operated. The on-shift

licensed operators will continue to be

supervised by personnel who hold an SRO

license in accordance with 10 CFR 50.54(l).

Therefore, neither of the proposed changes

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of § 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room Location:

California Polytechnic State

University, Robert E. Kennedy

Library, Government Documents and

Maps Department, San Luis Obispo,

California 93407

Attorney for Licensee: Christopher J.

Warner, Esq., Pacific Gas & Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25117

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: March

26, 1998.

Description of amendment request:

The proposed amendments would

revise Technical Specification (TS) 3/

4.8.2.1, ‘‘AC Distribution—Operating,’’

to add operability conditions and action

statements for the 115-volt vital

instrument bus (VIB) D and inverter.

The proposed amendments complete

the recommended action from NRC

Generic Letter 91–11, Resolution of

Generic Issues 48, ‘‘LCOs for Class 1E

Vital Instrument Buses,’’ and 49,

‘‘Interlocks and LCOs for Class 1E Tie

Breakers’’ pursuant to 10 CFR 50.54(f),

dated July 18, 1991.

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, as described above,

does not make any physical changes to the

plant or components, nor changes the

manner in which the plant or components

are operated as a result of the addition of the

Note and the D VIB and Inverter to the TS.

The proposed change incorporates the

operating requirements of the Technical

Specification Interpretation (TSI) developed

in response to GL 91–11 into the Salem Unit

1 and 2 Technical Specifications.

Incorporating this interpretation into the

Technical Specifications eliminates the need

for the TSI.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not introduce

any design or physical configuration change

to the plants, change the function of the 115

Volt D VIBs and inverters, or the manner in

which they are maintained or tested.

Therefore, the proposed amendment will

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 Action Times associated

with the incorporation of the D VIB into the

Technical Specifications are consistent with

the current Action Times for the A, B, and

C VIBs for a loss of an AC bus. Adding the

note to the Salem Unit 1 Technical

Specification brings consistency between

Salem Units 1 and 2, and is also consistent

with NUREG 1431, Vol. 1, Rev 1 ‘‘Standard

Technical Specifications Westinghouse

Plants.’’

The outage duration limit of 72 hours for

the D inverter is acceptable based on the

following: (1) the proposed 72 hours Action

Time to restore the inoperable inverter to

operable is supported by a PSA [probabilistic

safety assessment] assessment. NRC Draft

SRP [Standard Review Plan] Chapter 16.1,

Revision 13, ‘‘Risk-Informed Decision

making: Technical Specifications’’ notes that

an incremental conditional core damage

probability (ICCDP) of 5.0 E–7 is considered

very small. The proposed 72 hour allowable

outage time was calculated utilizing the NRC

incremental conditional core damage

probability (ICCDP), and (2) the inoperability

of the D VIB Inverter will not affect the

operation of any Safeguard Equipment

Cabinet (SEC) or Emergency Diesel Generator

(EDG).

Therefore, the proposed amendment will

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038

NRC Project Director: Robert A. Capra.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of amendment request: April 18,

1997, as supplemented by letters dated

October 10, 1997, and February 27,

1998.

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 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. The staff’s proposed

no significant hazards consideration

determination for the requested change

was published on June 4, 1997 (62 FR

30646).

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:

1a. Not 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 Limiting

Condition for Operation (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. Under

the proposed Action statements for

inoperable Station Vent Normal Range

Radiation Monitoring instrumentation,

should the control room normal ventilation

system be isolated and at least one train of

the control room emergency ventilation

system be placed in operation, these systems

would be in a state equivalent to that which

they would be in following an actual high

radiation condition. 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.

Not 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

Action statements for inoperable Station Vent

Normal Range Radiation Monitoring

instrumentation, should the control room

normal ventilation system be isolated and at

least one train of the control room emergency

ventilation system be placed in operation,

these systems would be in a state equivalent

to that which they would be in following an

actual high radiation condition. 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 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 consequences

of an accident.

2. Not create the possibility of a new or

different kind of accident from any accident
25118 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

previously evaluated because no new

accident initiators or assumptions are

introduced by the proposed changes.

As described above, under the proposed

Action statements for inoperable Station Vent

Normal Range Radiation Monitoring

instrumentation, should the control room

normal ventilation system be isolated and at

least one train of the control room emergency

ventilation system be placed in operation,

these systems would be in a state equivalent

to that which they would be in following an

actual high radiation condition. 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 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 would not introduce the possibility

of any new or different kinds of accidents.

3. Not 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 seven

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 one 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 one hour, but does not

require the CREVS be placed in operation.

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 would not affect 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:

University of Toledo, William Carlson

Library, Government Documents

Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Acting Project Director: Richard P.

Savio

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: March 9,

1998.

Description of amendment request:

The proposed amendment application

would revise Technical Specification

3/4.5.2b.1 and its associated Bases to

add clarification in regard to venting the

emergency core cooling system (ECCS)

pump casings and accessible discharge

piping high points. Technical

Specification 3/4.5.2b.1 requires

verification that the ECCS piping is full

of water at least once per 31 days by

venting the ECCS pump casings, i.e., the

safety injection pump, residual heat

removal pump, and centrifugal charging

pump casings and accessible discharge

piping high points. The centrifugal

charging pump (CCP) casings do not

have installed casing vents. Instead of a

casing vent, the suction and discharge

piping is installed as vertical runs

attached to the top-mounted suction and

discharge nozzles of each CCP pump.

Information provided by the pump

manufacturer indicates that the vertical

configuration of the piping is sufficient

to prevent the accumulation of

noncondensible gases that could cause

gas binding. Therefore the CCP casings

are effectively vented by vents on the

CCP discharge lines. The proposed

amendment application would revise

Technical Specification 3/4.5.2b.1 and

associated Bases to require the residual

heat removal and safety injection pump

casings and accessible ECCS discharge

piping high points be vented to ensure

the ECCS piping is full of water.

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 align the

surveillance requirements with the installed

system design and normal operating

conditions. The performance of surveillances

required by Technical Specifications is not

postulated to initiate an accident. The intent

of the surveillance ensures OPERABILITY of

the ECCS by verifying that the ECCS piping

is full of water and not subjected to gas

binding or water hammer. The design of the

CCPs is such that significant noncondensible

gases do not collect in the pumps, whether

they are running or not. Therefore, it is

unnecessary to require periodic pump casing

venting to ensure the CCPs will remain

OPERABLE. In addition, operating

experience has shown that no significant

voiding has occurred in the affected piping

which will continue to be vented at a high

point every 31 days per Surveillance

Requirement 4.5.2b.1). Therefore, no increase

in the probability or consequences of an

accident will occur as a result of this change.

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 will not result in

new failure modes because there are no

hardware changes nor are there any changes

in the method by which any safety-related

plant system performs its safety function.

The design of the CCPs is such that

significant noncondensible gases do not

collect in the pumps, whether they are

running or not. Therefore, it is not necessary

to require periodic pump casing venting to

ensure the equipment will remain

OPERABLE. Manual venting operations will

be performed to minimize the potential for

voids in system piping. Accordingly, this

change will not create the possibility of a

new or different kind of accident.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not affect the

acceptance criteria for any analyzed event.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems necessary to

assure the accomplishment of protective

functions. There will be no impact on any

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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 Missouri-Columbia,

Elmer Ellis Library, Columbia,

Missouri 65201–5149

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: William H.

Bateman

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request:

December 18, 1997.

Description of amendment request:

The proposed changes revise the

Technical Specifications (TS) to clarify

the terminology used to describe

equipment surveillances performed

with a refueling interval frequency.

Currently the TS are somewhat

ambiguous in the wording in this

regard, and the proposed changes would

adhere to the improved Standard TS
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25119

and make it clear whether the reactor

must be shutdown when performing the

test, or whether a ‘‘refueling interval’’

frequency (e.g., 18 months) is intended.

All of the clarifications are in Section 4

of the TS. In addition, minor

typographical errors are being corrected,

and an obsolete reference is proposed to

be deleted.

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—Operation of Surry Units 1

and 2 in accordance with the proposed

Technical Specifications change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The probability of an accident is not

increased as a result of the proposed

Technical Specification change since

surveillance intervals are being clarified, not

changed, and will continue to validate

system/component availability, operability

and performance during the appropriate unit

mode. The proposed change is administrative

in nature, therefore, station operations are

not being affected. The consequences of an

accident previously evaluated are not

increased since station operations are not

being changed, and no physical

modifications are being made to plant

systems or components.

Criterion 2—The proposed Technical

Specifications change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

As noted above, the proposed change is

administrative in nature. A new or different

type of accident is not being created since no

new accident precursors are being introduced

and equipment surveillances will continue to

be performed as required to ensure proper

system/component operation. Plant systems

are not being modified, system operations are

not being affected, and equipment

surveillance intervals are not being

increased. Consequently, the proposed

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

Criterion 3—The proposed Technical

Specifications change does not involve a

significant reduction in a margin of safety.

This is an administrative change.

Clarification of refueling surveillance interval

terminology to ensure consistency in

application does not affect plant equipment

performance. Surveillance intervals are not

being increased, and equipment surveillance

tests performed on a refueling interval

frequency (i.e. once per 18 months) will

continue to ensure system/component

performance as assumed in the existing

safety analyses. Therefore, the proposed

Technical Specification change does 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:

Swem Library, College of William and

Mary, Williamsburg, Virginia 23185

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia

23219

NRC Project Director: P.T. Kuo, Acting

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: March

25, 1998.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications (TS)

Sections 6.1.A; 6.1.A.2; 6.1.C.1.a and b;

6.1.C.1.f.1,4 and 8; 6.1.C.1.g.1 and 3;

6.8.A.2; and 6.8.B.2 for Units 1 and 2,

changing the title of Station Manager to

Site Vice President, and the titles of the

Assistant Station Managers to ManagerStation Operations and Maintenance

and Manager-Station Safety and

Licensing.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Virginia Electric and Power Company has

reviewed the proposed Technical

Specifications changes against the criteria of

10 CFR 50.92 and has concluded that the

changes do not pose a significant hazards

consideration. Specifically, station

operations in accordance with the proposed

Technical Specifications changes will not:

a. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes are administrative

in nature. The overall responsibility for safe

operation and review of plant operations is

not being changed. There are no changes to

the operation of any plant system or its

design as a result of these changes. Therefore,

neither the probability of occurrence nor the

consequences of an accident or malfunction

of equipment important to safety previously

evaluated in the safety analysis report are

increased.

b. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes are administrative

in nature. The overall responsibility for safe

operation and review of plant operations is

not being changed. There are no changes to

the operation of any plant system or its

design that could create any new modes of

operation or accident precursors. Therefore,

it is concluded that no new or different kind

of accident or malfunction from any

previously evaluated has been created.

c. The proposed changes do not result in

a significant reduction in margin of safety as

defined in the basis for any Technical

Specifications.

The proposed changes are administrative

in nature. The overall responsibility for safe

operation and review is not being changed.

There are no changes to the operation of any

plant system or its design as a result of these

changes. Safety systems are maintained

operable as required by Technical

Specifications. Therefore, the margin of

safety is not changed.

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:

Swem Library, College of William and

Mary, Williamsburg, Virginia 23185

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia

23219

NRC Project Director: P.T. Kuo, Acting

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: April 8,

1998.

Description of amendment request:

The change would reduce allowable

reactor coolant system (RCS) specific

activity from 1.0 microcurie/gram to

0.35 microcurie/gram dose equivalent

I–131.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change was reviewed in

accordance with the provisions of 10 CFR

50.92 to show no significant hazards exist.

The proposed change will not:

(1) Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The change implements a more restrictive

RCS activity limit. Specific RCS activity is an

initial plant condition and, therefore, is not

an accident initiator and can not cause the

occurrence of or increase the probability of

an accident. The change also lowers the

curve of Figure TS 3.1–3 which restricts

operation with high specific activity. The

new value for specific activity is justified by
25120 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

the Westinghouse calculation which

demonstrates acceptable offsite and control

room doses following a (main steamline

break) MSLB with a maximum allowable

primary to secondary leak rate. By lowering

the RCS specific activity and maintaining

leakage within the projected maximum

allowable, 10 CFR 100 and GDC 19 criteria

are satisfied. Therefore, the change does 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

previously evaluated.

The proposed change to the RCS specific

activity limit will not significantly effect

operation of the plant nor will it alter the

configuration of the plant. There will be no

additional challenges to the main steam

system or the reactor coolant system pressure

boundary and no new failure modes are

introduced. Therefore, the proposed change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Involve a significant reduction in the

margin of safety.

Reduction of the RCS specific activity limit

allows an increase in the MSLB allowable

primary to secondary leakage. The net effect

is no reduction in the margin of safety

provided by 10 CFR part 100 and GDC 19

criteria. The maximum allowable leakage is

the leakage limit for projected SG leakage

following SG tube inspection and repair.

Reducing specific activity to increase

projected leak rate follows guidance given by

GL 95–05 and effectively takes margin

available in the specific activity limits and

applies it to the projected SG leak rate. This

has been determined to be an acceptable

means for accepting higher projected leak

rates while still meeting the applicable limits

of 10 CFR part 100 and GDC 19 criteria with

respect to offsite and control room doses.

Additionally, monitoring of the specific

activity and compliance with the required

actions remains unchanged. Therefore, the

proposed change does not involve a

significant reduction in the margin of safety.

For consistency, the value of secondary

coolant activity in Table TS 4.1.2 is being

corrected from 1.0 microcurie/gram to 0.1

microcurie/gram. This is consistent with a

previously submitted and approved

amendment, therefore, no significant hazards

exist for this change.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room location:

University of Wisconsin, Cofrin

Library, 2420 Nicolet Drive, Green

Bay, WI 54311–7001

Attorney for licensee: Bradley D.

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

Box 1497, Madison, WI 53701–1497

NRC Project Director: Richard P. Savio

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: April 15,

1998.

Description of amendment request:

The revisions in the proposed Technical

Specification amendment are part of the

licensee’s fuel and reload change plan

for Cycle 23. The revisions implement

changes associated with a new fuel

design and also reflect changing plant

conditions due to steam generator tube

plugging and repair. The Technical

Specifications (TS) would be modified

as follows:

(1) Figure 2.1–1 would be revised to

reflect the recently approved High

Thermal Performance (HTP) Critical

Heat Flux (CHF) correlation and

corresponding Departure from Nucleate

Boiling Ratio (DNBR) limit of 1.14. The

figure would also reflect changes in

peak rod power and minimum reactor

coolant flow.

(2) TS 3.10.b—new hot channel

factors would be incorporated for the

new fuel design and the corresponding

increase in peaking factors. The limits

for Height Dependent Nuclear flux Hot

Channel Factor are specified in TS

3.10.b.1 and the limits for Nuclear

Enthalpy Rise Hot Channel Factor are

specified in 3.10.b.2.

(3) TS 3.10.k—the specification for

the maximum Reactor Coolant System

(RCS) Inlet Temperature would be

replaced with a specification for the

maximum Reactor Coolant System

(RCS) Average Temperature.

(4) TS 3.10.l—the statement ‘‘During

100% steady-state power operation’’

would be revised in the specification for

minimum Reactor Coolant System (RCS)

pressure and replaced with ‘‘During

steady-state power operation.’’

(5) TS 3.10.m—the minimum Reactor

Coolant Flow is being decreased to

85,500 gallons per minute per loop.

(6) TS 3.10.n—would be revised to

reflect the new Minimum DNBR limit.

(7) Figure TS 3.10–1—the Required

Shutdown Reactivity vs. Boron

Concentration would be revised to

reflect the change to an 18 month fuel

cycle.

(8) Figure TS 3.10–2, the Hot Channel

Factor Normalized Operating Envelope

would be revised to reflect the values

used in the new safety analyses.

(9) The Table of Contents and the

Basis sections would be revised to

accommodate the above 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:

Figure TS 2.1–1: The proposed changes

will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The safety limits curves are not accident

initiators. Therefore, the change will not

increase the probability of an accident

previously evaluated. The proposed changes

to the safety limits curves do not alter the

plant configuration, operating set points, or

overall plant performance. The safety limits

curves reflect the changes to the DNBR limit,

CHF correlation, RCS flow peaking factors

and fuel design. The significant hazards

determinations for these parameters are

evaluated later in this submittal. Therefore,

the change will not increase the

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 in the safety limits

curves do not alter the plant configuration,

operating set points, or overall plant

performance. Therefore, it does not create the

possibility of a new or different kind of

accident.

3. Involve a significant reduction in the

margin of safety.

Operation in the acceptable regions (i.e.,

below and to the left of the safety limit

curves) in combination with the reactor

protection and engineered safety systems

designed into the plant will ensure that the

safety limits are not exceeded during normal

operation or during anticipated design basis

operational transients. The core will be

operated in the nucleate boiling heat transfer

regime. Departure from nucleate boiling

(DNB) will not occur and therefore fuel

cladding integrity will be assured.

The revised safety limit curves have been

developed using operating parameters at

their bounding values (e.g., rod powers at the

peaking factor limits, reactor coolant flow at

the minimum operating limit). The revised

curves will bound plant operation with

Siemens Power Corporation standard or

heavy fuel. Therefore, this change will not

involve a significant reduction in safety

margin.

TS 3.10.b: The proposed changes will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Peaking factor limits are input assumptions

to the safety analyses and are not accident

initiators. Therefore, this change would not

increase the probability of occurrence of an

accident previously evaluated.

The safety analyses input assumptions are

designed to bound actual plant operation.

Changing the safety analysis input

assumption for the increased peaking factor

limits does not change the underlying

progression of design basis accidents

evaluated in the safety analyses. All safety

analysis acceptance criteria are satisfied in

the increased peaking factor limit conditions.

Additionally, the radiological consequences
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25121

are bounded by existing analysis at the

increased peaking factor limits. Therefore,

this change will not significantly increase the

consequences of an accident previously

analyzed.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This change incorporates the safety

analyses assumptions for core peaking factor

limits for Siemens Power Corporation heavy

fuel. The change does not alter plant

equipment, set points or plant performance.

Therefore, changing the peaking factor limits

for analysis purposes will not create a new

or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in the

margin of safety.

Results of the safety analyses and of

radiological consequences indicate that all

acceptance criteria are satisfied. The peaking

factor limits assumed in the safety analyses

are consistent with the proposed revised

limits and these revised limits are established

to bound actual plant operation. Therefore,

this change will not involve a significant

reduction in the margin of safety.

TS 3.10.k: The proposed change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The RCS average temperature limit is not

an accident initiator. Changing the technical

specification limit consistent with the

accident analyses will not increase the

probability of an accident previously

evaluated.

The proposed change limits the maximum

reactor coolant system average temperature to

568.8 °F. The design basis safety analyses,

the Large and Small Break LOCA accidents

and the non-LOCA accidents, have been

analyzed and/or evaluated consistent with

the revised RCS average temperature. The reanalysis and evaluation have demonstrated

that all safety analysis acceptance criteria are

satisfied at the specified temperature.

Therefore, the change will not increase the

consequences of an accident previously

evaluated.

The proposed technical specification limit

for maximum allowed RCS average

temperature was decreased below the

analytical limit to account for instrument

error.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not alter the

plant configuration, operating set points, or

overall plant performance. Therefore, it does

not create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in the

margin of safety.

The proposed change is consistent with the

safety analyses. All safety analyses

acceptance criteria are satisfied at the revised

reactor coolant system average temperature.

The TS limit will bound actual plant

operation. Therefore, there is no significant

reduction in the margin of safety.

TS 3.10.l: The proposed change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The RCS pressure limit is not an accident

initiator. By removing the 100% value from

the specification, the assumptions in the

safety analyses are not changed. Changing the

technical specification to remove the 100%

power criteria will not increase the

probability of an accident previously

evaluated.

The design basis safety analyses have been

analyzed and/or evaluated at the specified

RCS pressure. The analyses and evaluations

have demonstrated that all safety analyses

acceptance criteria are satisfied at this

pressure. Therefore, the change would not

increase the consequences of an accident

previously evaluated.

The proposed technical specification limit

for minimum allowed RCS pressure was

increased above the analytical limit to

account for instrument error.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not alter the

plant configuration, operating set points, or

overall plant performance. Therefore, it does

not create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in the

margin of safety.

The proposed change is consistent with the

safety analyses. All safety analyses

acceptance criteria are satisfied at the reactor

coolant system pressure. The limit will

bound actual plant operation. Therefore,

there is no significant reduction in the

margin of safety.

TS 3.10.m: The proposed change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The RCS flow limit is not an accident

initiator. Changing the technical specification

limit consistent with the accident analysis

will not increase the probability of an

accident previously evaluated.

The proposed change limits the minimum

reactor coolant flow. The design basis safety

analyses have been analyzed and/or

evaluated at the revised RCS flow. The reanalysis and evaluation have demonstrated

that all safety analysis acceptance criteria are

satisfied at the specified flow. Therefore, the

change will not significantly increase the

consequences of an accident previously

evaluated.

The proposed technical specification limit

for minimum allowed RCS flow was

increased above the analytical limit to

account for instrument error.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not alter the

plant configuration or overall plant

performance. Therefore, it does not create the

possibility of a new or different kind of

accident.

3. Involve a significant reduction in the

margin of safety.

The proposed change is consistent with the

safety analyses. All safety analyses

acceptance criteria are satisfied at the revised

reactor coolant system flow. The limit will

bound actual plant operation.

The change reduces the RCS flow rate

limit. Re-analysis of LOCA and non-LOCA

transients determined all safety requirements

of KNPP accident analyses were still met at

the reduced RCS flow rate limit. Therefore,

this proposed change does not significantly

reduce the margin of safety.

TS 3.10.n: The proposed change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The Departure from Nucleate Boiling Ratio

(DNBR) is not an accident initiator.

Therefore, the change in the DNBR will not

increase the probability of an accident

previously evaluated.

The proposed change to the DNBR value

does not change plant configuration,

operating set points, or overall plant

performance. Therefore, the change will not

increase the 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 change does not alter the

plant configuration, operating set points, or

overall plant performance. Therefore, it does

not create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in the

margin of safety.

All safety analyses acceptance criteria are

satisfied using the HTP CHF correlation. The

DNBR limits assumed in the safety analyses

will bound actual plant operation and

assures at 95/95 that DNBR will not occur.

Therefore, there is no reduction in the margin

of safety.

TS Figure 3.10–1: The proposed change

will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Required Shutdown Reactivity vs. Boron

Concentration was revised to reflect the

longer cycle length and the resulting increase

in boron concentration. The Required

Shutdown Reactivity vs. Boron

Concentration is not an accident initiator.

Extending the boron concentrations to

account for longer fuel cycles will not

increase 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 change does not alter the

plant configuration, operating set points, or

overall plant performance. Therefore, it does

not create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in the

margin of safety.

The proposed change is consistent with the

cycle length and core physics analyses for

longer fuel cycles. Operation within the

limits specified in the figure will assure all

core safety evaluation acceptance criteria are

satisfied. The limit will bound actual plant

operation. Therefore, there is no reduction in

the margin of safety.

TS Figure 3.10–2: The proposed change

will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.
25122 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

The Hot Channel Factor Normalized

Operating Envelope figure was revised to

reflect the values used in the safety analyses.

The Hot Channel Factor Normalized

Operating Envelope figure is not an accident

initiator. Changing the technical specification

figure consistent with the assumptions of the

accident analyses will not increase 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 change does not alter the

plant configuration, operating set points, or

overall plant performance. Therefore, it does

not create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in the

margin of safety.

The proposed change is consistent with the

safety analyses. Operation within the limits

specified in the figure will assure all safety

analyses acceptance criteria are satisfied. The

limit will bound actual plant operation.

Therefore, there is no 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:

University of Wisconsin, Cofrin

Library, 2420 Nicolet Drive, Green

Bay, WI 54311–7001

Attorney for licensee: Bradley D.

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

Box 1497, Madison, WI 53701–1497

NRC Project Director: Richard P. Savio

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Units 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: May 2,

1995, October 12, 1995, March 26, 1996,

and December 15, 1997 (TSCR 172)

Description of amendment request:

The proposed amendments would

revise Technical Specifications (TS)

Table 15.4.1–1, ‘‘Minimum Frequencies

for Checks, Calibrations, and Tests of

Instrument Channels,’’ to change the

test frequencies for radiation monitors

as discussed in Generic Letter 93–05

(‘‘Line-Item Technical Specifications

Improvements To Reduce Surveillance

Requirements For Testing During Power

Operation’’), remove the radiation

monitoring system as item 36, revise

note(s), and add those radiation

monitors and their surveillance

requirements that support current TS or

meet the requirements of 10 CFR 50.36.

Additionally, several typographical and

nomenclature errors would be corrected.

This amendment request was initially

noticed in the Federal Register on June

6, 1995 (60 FR 29890).

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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

1. Operation of this facility under the

proposed TS will not create a significant

increase in the probability or

consequences of an accident previously

evaluated.

The probabilities of accidents

previously evaluated are based on the

probability of initiating events for these

accidents. Initiating events for accidents

previously evaluated for the Point Beach

Nuclear Plant (PBNP) include control

rod withdrawal and drop, chemical

volume control system malfunction

(boron dilution), startup of an inactive

reactor coolant loop, reduction in

feedwater enthalpy, excessive load

increase, losses of reactor coolant flow,

loss of external electrical load, loss of

normal feedwater, loss of all alternating

current (ac) power to the auxiliaries,

turbine overspeed, fuel handling

accidents, accidental releases of waste

liquid or gas, steam generator tube

rupture, steam pipe rupture, control rod

ejection, and primary coolant system

ruptures.

These proposed changes do not cause

an increase in the probabilities of any

accidents previously evaluated because

these changes will not cause an increase

in the probability of any initiating

events for accidents previously

evaluated. In particular, these changes

affect the radiation monitoring system

surveillance requirements and make

administrative changes that will not

result in changing accident initiators.

The consequences of the accidents

previously evaluated in the Final Safety

Analysis Report (FSAR) are determined

by the results of analyses that are based

on initial conditions of the plant, the

type of accident, transient response of

the plant, and the operation and failure

of equipment and systems.

The proposed changes reduce the

burden associated with radiation

monitoring system required surveillance

by establishing surveillances for only

the necessary monitors (i.e., elimination

of the testing requirement for monitors

that do not perform a required function)

and changing the testing frequency for

these monitors from monthly to

quarterly. The proposed changes do not

increase the probability of failure of this

equipment or its ability to operate as

required for the accidents previously

evaluated in the PBNP FSAR. The

proposed changes to correct

typographical errors and correct

nomenclature are administrative only

and do not increase the probability of an

accident previously evaluated nor do

they affect the consequences of any

accident previously evaluated.

Therefore, these proposed license

amendments do not affect the

consequences of any accident

previously evaluated in the PBNP FSAR

because the factors that are used to

determine consequences of accidents

are not being changed.

2. Operation of this facility under the

proposed TS change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

New or different kinds of accidents

can only be created by new or different

accident initiators or sequences. The

changes proposed by this license

amendment request do not create any

new or different accident initiators or

sequences because the revisions to TS

Table 15.4.1–1, ‘‘Minimum Frequencies

for Checks, Calibrations, and Tests of

Instrument Channels,’’ will not cause

failures of equipment or accident

sequences different than the accidents

previously evaluated. The proposed

changes to correct typographical errors

and correct nomenclature are

administrative only. Therefore, these

proposed TS changes do not create the

possibility of an accident of a different

type than any previously evaluated in

the Point Beach FSAR.

3. Operation of this facility under the

proposed TS change will not create a

significant reduction in a margin of

safety.

The margins of safety for Point Beach

are based on the design and operation

of the reactor and containment and the

safety systems that provide their

protection. The changes proposed by

this license amendment request provide

the appropriate surveillance

requirements for the radiation

monitoring system. The revised

surveillance requirements will continue

to ensure that the required radiation

monitors will operate as required. The

design and operation of the reactor and

containment are not affected by these

proposed changes. The proposed

changes to correct typographical errors

and correct nomenclature are

administrative only. Therefore, the

margins of safety for Point Beach are not

being reduced because the design and

operation of the reactor and

containment are not being changed.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25123

proposes to determine that the

amendment request involves no

significant hazards considerations.

Local Public Document Room location:

The Lester Public Library, 1001

Adams Street, Two Rivers, Wisconsin

54241

Attorney for licensee: John H. O’Neill,

Jr., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Cynthia A.

Carpenter

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Carolina Power & Light Company, et al.,

Docket No. 50–325, Brunswick Steam

Electric Plant, Unit 1, Brunswick

County, North Carolina

Date of amendment request: February

23, 1998, as supplemented March 27,

1998.

Brief description of amendment: The

proposed amendment would allow

addition of a footnote to the Safety Limit

Minimum Critical Power Ratio value in

the Technical Specifications and the

associated action statement.

Date of publication of individual

notice in the Federal Register: April 10,

1998 (63 FR 17900).

Expiration date of individual notice:

May 11, 1998.

Local Public Document Room location:

University of North Carolina at

Wilmington, William Madison

Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: April 3,

1998, and related application dated

November 22, 1995, as supplemented

February 19, April 19, May 3, June 12,

and December 4, 1996, and January 30

and August 7, 1997.

Description of amendment request:

The proposed amendment would revise

Technical Specification 3.8.1.1 to

change the emergency diesel generator

allowed outage time from 3 to 7 days.

This would be a one-time amendment,

effective from the date of issuance until

September 30, 1998.

Date of publication of individual

notice in Federal Register: April 13,

1998 (63 FR 18048).

Expiration date of individual notice:

May 13, 1998.

Local Public Document Room location:

Monroe County Library System, 3700

South Custer Road, Monroe, Michigan

48161

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request: April 9,

1998, TXX–98107.

Description of amendment request:

The proposed amendment would allow

on a one time basis, the verification of

the proper operation of the Unit 2 load

shed seal-in contacts and the diesel

generator trip bypass contacts at power

and crediting performance of

Surveillance Requirements (SR)

4.8.1.1.2f.4(a) and 4.8.1.1.2f.6(a), at

power as opposed to ‘‘during

shutdown’’ as currently required by

those SR. The proposed amendment

would also allow on a one time basis the

verification of the proper operation of

the Unit 2 lockout relays and contacts

to be deferred until the startup from

2RFO4 or earlier outage to at least

MODE 3.

Date of individual notice in the

Federal Register: April 20, 1998.

Expiration date of individual notice:

May 5, 1998.

Local Public Document Room location:

University of Texas at Arlington

Library, Government Publications/

Maps, 702 College, P.O. Box 19497,

Arlington, TX 76019

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 Ch. 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.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

March 17, 1997, as supplemented April

13, 1998. The April 13, 1998, submittal

contained clarifying information only,

and did not change the proposed no

significant hazards consideration.

Brief description of amendment: The

amendment revises Technical

Specifications 4.1.2.2.c, 4.5.2.e,

4.6.2.1.c, 4.6.2.2.c, 4.6.3.2, 4.7.1.2.1.b,

4.7.3.b, and 4.7.4.b to delete specific

restrictions in the text of the

surveillances that the tests must be done

while the unit is shut down.

Date of issuance: April 14, 1998.

Effective date: April 14, 1998

Amendment No.: 77.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: April 23, 1997 (62 FR 19826)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 14, 1998.
25124 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Cameron Village Regional Library,

1930 Clark Avenue, Raleigh, North

Carolina 27605

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

December 12, 1997.

Brief description of amendments: The

amendments modify the bypass logic for

Main Steam Line Isolation Valve

Isolation Actuation Instrumentation on

Condenser Low Vacuum as stated in

Technical Specification Tables 3.3.2–1

and 4.3.2–1.

Date of issuance: April 14, 1998.

Effective date: Immediately, to be

implemented prior to startup from

L1F35 for Unit 1 and from L2R07 for

Unit 2.

Amendment Nos.: 124 and 109.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 11, 1998 (63 FR

6982).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 14, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Jacobs Memorial Library, Illinois

Valley Community College, Oglesby,

Illinois 61348

Duke Energy Corporation, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

December 18, 1997, as supplemented by

letter dated January 26, 1998.

Brief description of amendments: The

amendments revise the operating

license of Unit 1 and Unit 2 to (1) delete

license conditions that have been

fulfilled; (2) delete exemptions that have

expired; (3) update information to

reflect current plant status and

regulatory requirements; and (4) make

other corrections and editorial changes.

Date of issuance: April 23, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1–164; Unit

2–156.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Operating Licenses.

Date of initial notice in Federal

Register: February 11, 1998 (63 FR

6983).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 23, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

York County Library, 138 East Black

Street, Rock Hill, South Carolina

Duke Energy Corporation, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

March 3, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications to change the

qualification requirements for the

members of the Safety Review Group.

Date of issuance: April 27, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1–165; Unit

2–157.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: March 25, 1998 (63 FR 14486).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 27, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

York County Library, 138 East Black

Street, Rock Hill, South Carolina.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

August 28, 1997. Supplement January

22, February 19, March 19, and April 6,

13, and 17, 1998.

Brief description of amendments: The

amendments incorporate new testing

and operability requirements related to

the installation of new systems and

upgrades associated with the Emergency

Condenser Circulating Water System.

Review of the system for this

amendment also includes a review of

the new design features incorporated

into the upgrade and its acceptability as

a safety grade system.

Date of Issuance: April 24, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1–229; Unit

2–230; Unit 3–226

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: The

amendments revised the Technical

Specifications and Appendix C of the

Operating Licenses.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50002).

The January 22, 1998, February 19,

March 19, and April 6, 13, and 17, 1998,

letters provided clarifying information

that did not change the scope of the

August 28, 1997, application and the

initial proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated April 24, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Oconee County Library, 501 West

South Broad Street, Walhalla, South

Carolina

Duquesne Light Company, et al., Docket

Nos. 50–334 and 50–412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

March 16, 1998.

Brief description of amendments:

These amendments add a new Limiting

Condition for Operation (LCO) 3.0.6 to

TS Section 3/4.0, ‘‘APPLICABILITY.’’

The new LCO 3.0.6 provides specific

guidance for returning equipment to

service under administrative control to

perform testing required to demonstrate

OPERABILITY.

Date of issuance: April 15, 1998.

Effective date: Both units, effective

immediately, to be implemented within

30 days.

Amendment Nos.: 213 and 90.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (63 FR 14142, March

24, 1998). That notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by April 23, 1998,

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 amendments, finding of exigent

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated April 15, 1998.
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25125

Local Public Document Room location:

B.F. Jones Memorial Library, 663

Franklin Avenue, Aliquippa, PA

15001

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 26,

1997, as supplemented by letter dated

September 11, 1997.

Brief description of amendment: The

amendment changes the Appendix A

TSs by modifying Tables 3.7–1 and 3.7–

2. The revision to Table 3.7–1 changes

the Main Steam Safety Valves (MSSVs)

orifice size from 26 square inches to

28.27 square inches and relocates the

orifice size from the TS Table to the TS

Bases. The change to correct the orifice

size is an editorial change to make the

TS consistent with plant design. The

changes to Table 3.7–2 delete the

provisions that allows continued plant

operation with three MSSVs inoperable.

The proposed amendment will also

revise TS Bases 3/4.7.1.1 to remove the

equation used for determining the

reduced maximum allowable linear

power level-high reactor trip settings of

TS Table 3.7–2.

Date of issuance: April 20, 1998.

Effective date: April 20, 1998, to be

implemented within 30 days.

Amendment No.: 142.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38135).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 20, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

University of New Orleans Library,

Louisiana Collection, Lakefront, New

Orleans, LA 70122

GPU Nuclear, Inc. and Saxton Nuclear

Experimental Corporation (SNEC),

Docket No. 50–146, Saxton Nuclear

Experimental Facility (SNEF)

Date of application for amendment:

November 25, 1996, as supplemented on

May 30, June 4 and 16, August 21 and

September 16, 1997, and February 3 and

9, 1998, and March 31, 1998. During the

amendment request review, the staff

also referred to the SNEF

Decommissioning Environmental Report

dated April 17, 1996, licensee responses

to NRC questions about the

environmental report dated July 18,

1996, and March 3 and 31, 1998, the

SNEC Facility Updated Safety Analysis

Report, Revision 0, submitted on

October 25, 1996, Revision 1, submitted

on August 21, 1997, and Revision 2,

submitted on February 3, 1998, and the

SNEC Facility Decommissioning Quality

Assurance Plan submitted by letter

dated November 8, 1996, as

supplemented on May 30, 1997, and

February 3 and 9, 1998.

Brief description of amendment: The

amendment allows decommissioning of

the SNEF. The changes to the license

and Technical Specifications (TSs) (1)

accommodate decommissioning

activities at the SNEF, (2) establish

specific TS controls over

decommissioning activities, (3) establish

limiting conditions for performing

decommissioning activities, (4) extend

exclusion area controls to include the

SNEF Decommissioning Support

Facility, (5) establish requirements for a

Radiological Environmental Monitoring

Program, and an Offsite Dose

Calculation Manual, and (6) establish

requirements for technical and

independent safety reviews. In addition,

the amendment authorizes other

administrative and editorial changes to

the TSs associated with the changes

described above.

Date of issuance: April 20, 1998.

Effective date: April 20, 1998.

Amendment No.: 15.

Amended Facility License No. DPR–4:

Amendment changed the Amended

Facility License and TSs.

Date of initial notice in Federal

Register: March 12, 1997 (62 FR 11494).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 20, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room Location:

Saxton Community Library, Front

Street, Saxton, Pennsylvania 16678

GPU Nuclear Corporation, et al., Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit No. 1 (TMI–1), Dauphin

County, Pennsylvania

Date of application for amendment:

December 16, 1996, as supplemented

September 11, 1997 and March 25,

1998.

Brief description of amendment: The

amendment (1) reflects the change in

the legal name of the operator of TMI–

1 from GPU Nuclear Corporation to GPU

Nuclear, Inc., and (2) reflects in the

TMI–1 Facility Operating License the

registered trade name of GPU Energy

now used by the owners of the facility.

Date of Issuance: April 24, 1998.

Effective Date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 207.

Facility Operating License No. NPF–

50: Amendment revised the Facility

Operating License and the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4350).

The September 11, 1997 and March

25, 1998, submittals provided clarifying

information and did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 24, 1998.

No significant hazards consideration

comments received: No.

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

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit No. 2, Oswego

County, New York

Date of application for amendment:

October 7, 1997.

Brief description of amendment: The

amendment revised the Technical

Specifications surveillance

requirements to change setpoints for the

refueling platform main hoist overload

cutoff, loaded interlock, and redundant

loaded interlock due to planned

modifications to the refueling platform

mast.

Date of issuance: April 16, 1998.

Effective date: As of the date of

issuance to be implemented upon

completion and acceptance of design

modifications to the refueling platform

mast.

Amendment No.: 81.

Facility Operating License No. NMF–

69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68309).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 16, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Reference and Documents

Department, Penfield Library, State

University of New York, Oswego,

New York 13126

Northern States Power Company,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of application for amendment:

March 13, 1998, as supplemented March

25, 1998.
25126 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

Brief description of amendment: The

amendment modifies the Technical

Specification requirements associated

with the Minimum Critical Power Ratio

(MCPR) safety limits for Cycle 19 based

on the cycle-specific analysis of the

current mixed core of GE [General

Electric] 11, GE10, four GE12 lead use

assemblies, and eight SPC [Siemens

Power Corporation] ATRIUM–9B

assemblies.

Date of issuance: April 20, 1998.

Effective date: April 20, 1998.

Amendment No.: 100.

Facility Operating License No. DPR–

22: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 20, 1998 (63 FR 13704).

The March 25, 1998, letter provided

clarifying information in response to the

staff’s request for additional information

during a teleconference. This

information was within the scope of the

original application and did not change

the staff’s initial proposed no significant

hazards considerations determination.

Therefore, renoticing was not

warranted.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 20, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of application for amendment:

October 14, 1997, as supplemented on

March 26, 1998.

Brief description of amendment: The

amendment revises Technical

Specification (TS) 3.4.6.3, ‘‘Primary

Coolant System Pressure Isolation

Valves Limiting Condition for

Operation,’’ to add additional pressure

isolation valves, establish the

operability and testing requirements for

the pressure isolation valves, and make

this section more consistent with Salem

Unit 2 TSs.

Date of issuance: April 20, 1998.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 210.

Facility Operating License No. DPR–

70: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 19, 1997 (62 FR

61845).

The March 26, 1998, letter 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 April 20, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Salem Free Public Library, 112 West

Broadway, Salem, NJ 08079.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

January 26, 1998.

Brief description of amendments: The

proposed amendments would (1)

modify the requirement to hold a

Susquehanna Steam Electric Station

(SSES) Senior Reactor Operator (SRO)

license in Section 6.3.1 for the ManagerNuclear Operations (MNO), (2) replace

the position of MNO with Operations

Supervisor—Nuclear in the Section

6.2.2g requirement to hold an SSES SRO

license and (3) renumber existing TS

Section 6.3.1 to include 6.3.1.1, 6.3.1.2,

and 6.3.1.3.

Date of issuance: April 10, 1998.

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 175 and 147.

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 24, 1998 (63 FR

9270).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 10, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Osterhout Free Library, Reference

Department, 71 South Franklin Street,

Wilkes-Barre, PA 18701

Pennsylvania Power and Light

Company, Docket No. 50–388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of application for amendment:

January 11, 1996, as supplemented

March 16, 1998.

Brief description of amendment: This

amendment changes the TSs to preclude

the need to enter into Limiting

Condition for Operation 3.0.3 to allow

performance of certain emergency diesel

generator testing.

Date of issuance: April 10, 1998.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 148.

Facility Operating License No. NPF–

22: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10397).

The February 15, 1996, letter

corrected the no significant hazards

(NSH) determination. The NSH

determination was used in the March

13, 1996 (61 FR 10397) notice. The

March 24, 1998, letter 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 April 10, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Osterhout Free Library, Reference

Department, 71 South Franklin Street,

Wilkes-Barre, PA 18701

Philadelphia Electric Company, Docket

No. 50–352, Limerick Generating

Station, Unit 1, Montgomery County,

Pennsylvania

Date of application for amendment:

January 12, 1998.

Brief description of amendment: This

amendment revises TS Table 4.4.6.1.3–

1 to change the withdrawal schedule for

the first capsule to be withdrawn from

10 Effective Full Power Years (EFPY) to

15 EFPY. In addition, TS Surveillance

Requirement 4.4.6.1.4 will be revised to

remove the references to flux wire

removal and analysis that was originally

required following the first cycle of

operation and replaced with a new

surveillance requirement. The new

requirement refers to the flux wires that

are located within the surveillance

capsules, which will be removed and

analyzed in accordance with the

surveillance capsule removal schedule

located in Table 4.4.6.1.3–1.

Date of issuance: April 15, 1998.

Effective date: As of the date of

issuance.

Amendment No.: 126.

Facility Operating License No. NPF–

39: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 11, 1998 (63 FR

6988).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 15, 1998.

No significant hazards consideration

comments received: No.
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25127

Local Public Document Room location:

Pottstown Public Library, 500 High

Street, Pottstown, PA 19464

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

February 27, 1998.

Brief description of amendment: The

amendment changes the Technical

Specifications by revising the pressuretemperature curves to extend heatup

and cooldown limits from 11 to 13.3

effective full-power years, provides the

corresponding overpressure protection

system limits, and makes some minor

changes to ensure specification clarity

and conservatism.

Date of issuance: April 10, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 179.

Facility Operating License No. DPR–

64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 9, 1998 (63 FR 11456).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 10, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

White Plains Public Library, 100

Martine Avenue, White Plains, New

York 10610

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

February 26, 1998, as supplemented by

letter dated March 20, 1998.

Brief description of amendment: This

amendment revises Technical

Specification (TS) Section 3/4.4.5,

‘‘Reactor Coolant System—Steam

Generators,’’ TS Section 3/4.4.6.2,

‘‘Reactor Coolant System—Operational

Leakage,’’ and the associated bases to

allow use of the ‘‘repair roll’’ steam

generator tube repair process.

Date of issuance: April 14, 1998.

Effective date: April 14, 1998.

Amendment No.: 220.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 9, 1998 (63 FR 11460).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 14, 1998.

No significant hazards consideration

comments received: No. The

supplemental information submitted by

the licensees did not affect the proposed

no significant hazards consideration

determination.

Local Public Document Room location:

University of Toledo, William Carlson

Library, Government Documents

Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

June 24, 1997.

Brief description of amendment: This

amendment revises TS Section 3/

4.3.2.1, ‘‘Safety Features Actuation

System Instrumentation,’’ TS Section 3/

4.6.1.7, ‘‘Containment Ventilation

System,’’ TS Section 3/4.6.3.1,

‘‘Containment Isolation Valves,’’ and TS

Section 3/4.9.4, ‘‘Refueling

Operations—Containment

Penetrations,’’ and the associated TS

Bases. Valve position requirements have

been added, and certain containment

radiation monitor requirements, valve

isolation verification requirements, and

containment radiation monitor optional

uses have been deleted. Administrative

changes have also been made.

Date of issuance: April 15, 1998.

Effective date: April 15, 1998.

Amendment No.: 221.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40858).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 15, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

University of Toledo, William Carlson

Library, Government Documents

Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment requests:

February 25, 1998, (TXX–98050) as

supplemented by letter dated March 9,

1998, (TXX–98066) for License

Amendment Request (LAR) 98–002,

March 12, 1998, (TXX–98076) for LAR

98–003, and March 18, 1998, (TXX–

98079) for LAR 98–004.

Brief description of amendments: This

amendment is the result of three Notice

of Enforcement Discretions (NOEDs)

dated February 24, March 13, and 17,

1998. These NOEDs although distinct

actions changed the same page of the

CPSES TS therefore the single

amendment is being issued to cover the

three parts of this amendment.

The first part of the amendment

would be a temporary change to the TSs

to remove the requirement to

demonstrate the load shedding feature

of MCC XEB4–3 as part of Surveillance

Requirements (SRs) 4.8.1.1.2f.4)a) and

4.8.1.1.2f.6)a) until the plant startup

subsequent to the next refueling outage

for Unit or until an outage of 24 hour

in duration.

The second part of the amendment

would provide a temporary Technical

Specification change for SRs

4.8.1.1.2f.4)b) and 4.8.1.1.2f.6)b) to

allow the verification of the auto

connected shut-down loads through the

load sequencer to be performed at

power for fuel cycle 6 on Unit 1 and fuel

cycle 4 on Unit 2.

The third part of the amendment

would allow on a one time basis,

crediting performance of Surveillance

Requirements (SR) 4.8.1.1.2f.4)a) and

4.8.1.1.2f.6)a), during POWER

OPERATIONS as opposed to ‘‘during

shutdown.’’ Note that the bus tie breaker

for MCC XEB4–3 for Unit 2 was not

tested during the last surveillance test

and was the subject of part one of this

amendment.

Date of issuance: April 20, 1998.

Effective date: April 20, 1998.

Amendment Nos.: Unit 1—

Amendment No. 58; Unit 2—

Amendment No. 44.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 9, 1998 (63 FR 11458),

March 27, 1998 (63 FR 14974) and April

2, 1998 (63 FR 16287).

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated April 20, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

University of Texas at Arlington

Library, Government Publications/

Maps, 702 College, PO Box 19497,

Arlington, TX 76019

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

October 31, 1997, as supplemented by

letter dated February 27, 1998.

Brief description of amendment: The

amendment revises the Callaway Plant,
25128 Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices

Unit 1 Technical Specifications to

change setpoint and allowable stress

values of certain reactor trip system

(RTS) and engineered safety features

actuation system (ESFAS) functional

units.

Date of issuance: April 13, 1998.

Effective date: April 13, 1998, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 125.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 14, 1998 (63 FR 2283).

The February 27, 1998, supplemental

letter provided additional clarifying

information that did not change the

staff’s original no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 13, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

University of Missouri-Columbia,

Elmer Ellis Library, Columbia,

Missouri 65201–5149

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of application for amendment:

December 11, 1997, as supplemented on

March 3, 1998.

Brief description of amendment: The

amendment revises the values for the

safety limit minimum critical power

ratio for Cycle 20 operation.

Date of Issuance: April 10, 1998.

Effective date: April 10, 1998, to be

implemented within 30 days.

Amendment No.: 159.

Facility Operating License No.DPR–

28. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 11, 1998, (63 FR

7000).

The March 3,1998 supplement did not

change the original proposed no

significant hazards consideration.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated April 10, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Brooks Memorial Library, 224 Main

Street, Brattleboro, VT 05301

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of application for amendment:

September 11, 1996, as supplemented

by letter dated December 8, 1997.

Brief description of amendment: The

amendment involves a change to the

safety and relief valve setpoint tolerance

and power operation with an inoperable

safety relief valve.

Date of Issuance: April 15, 1998.

Effective date: April 15, 1998, to be

implemented within 30 days.

Amendment No.: 160.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17241).

The information provided in the

December 8, 1997, submittal did not

change the original proposed no

significant hazards determination.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated April 15, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

Brooks Memorial Library, 224 Main

Street, Brattleboro, VT 05301

Virginia Electric and Power Company, et

al., Docket Nos. 50–338 and 50–339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

November 26, 1996.

Brief description of amendments: The

proposed action would revise the

Technical Specifications (TS) to

eliminate the records retention

requirements from Section 6.10 of the

TS since these requirements have

already been relocated to the

Operational Quality Assurance program,

Chapter 17, in revision 32 of the

Updated Final Safety Analysis Report.

Date of issuance: April 13, 1998.

Effective date: April 13, 1998.

Amendment Nos.: 208 and 189.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 132).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 13, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498

Virginia Electric and Power Company, et

al., Docket Nos. 50–338 and 50–339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

February 3, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) Surveillance

Requirement Tables 3.3–1 and 4.3–1 for

both units, modifying the testing

requirements for the reactor trip bypass

breaker.

Date of issuance: April 14, 1998.

Effective date: April 14, 1998.

Amendment Nos.: 209 and 190.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 11, 1998 (63 FR 11925).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 14, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498

Virginia Electric and Power Company, et

al., Docket Nos. 50–338 and 50–339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

November 18, 1997.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) Surveillance

Requirements 4.7.1.7.2.a.1 and

4.7.1.7.2.a.2 for both units, modifying

the testing frequency of the Turbine

throttle and Governor valves.

Date of issuance: April 16, 1998.

Effective date: April 16, 1998.

Amendment Nos.: 210 and 191.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 17, 1997 (62 FR

66146)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 16, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498
Federal Register / Vol. 63, No. 87 /Wednesday, May 6, 1998 / Notices 25129

Virginia Electric and Power Company, et

al., Docket Nos. 50–338 and 50–339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

February 3, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) Surveillance

Requirement 4.4.10.1.1, modifying the

inspection requirements for the Reactor

Coolant Pump (RCP) flywheels for both

units and eliminating the examination

requirements for the flow straighteners

in each steam generator to the RCP

elbow on Unit 1.

Date of issuance: April 22, 1998.

Effective date: April 22, 1998.

Amendment Nos.: 211 and 192.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 11, 1998 (63 FR 11924)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 22, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room location:

The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498

Dated at Rockville, Md., this 29th day of

April 1998.

For the Nuclear Regulatory Commission.

Stuart A. Richards,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–11911 Filed 5–5–98; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

to Withdraw from Listing and

Registration; (Oryx Technology Corp.,

Common Stock, $0.001 Par Value;

Common Stock Warrants) File No. 1–

12680

April 30, 1998.

Oryx Technology Corp. (‘‘Company’’)

has filed an application with the

Securities and Exchange Commission

(‘‘Commission’’), pursuant to Section

12(d) of the Securities Exchange Act of

1934 (‘‘Act’’) and Rule 12d2–2(d)

promulgated thereunder, to withdraw

the above specified securities

(‘‘Securities’’) from listing and

registration on the Pacific Exchange,

Inc. (‘‘PCX’’ or ‘‘Exchange’’).

The reasons cited in the application

for withdrawing the Securities from

listing and registration include the

following:

The Securities of the Company have

been listed for trading on the Exchange

and, pursuant to a Registration

Statement of Form 8–A, effective on

April 5, 1994, the National Association

of Securities Dealers Automated

Quotation System (‘‘NASDAQ’’).

Trading in the Company’s Securities on

the NASDAQ commenced at the

opening of business on April 6, 1994,

and concurrently therewith on the PCX.

The Company has complied with

Exchange Rule 3.4(b) by filing with the

Exchange a certified copy of the

resolutions adopted by the Company’s

Board of Directors authorizing the

withdrawal of the Securities from listing

and registration on the PCX and by

setting forth in detail to the Exchange

the reasons for and facts supporting the

proposed delisting. In deciding to

withdraw its Securities from listing and

registration of the PCX, the Company

considered the direct and indirect costs

and expenses attendant on maintaining

the dual listing of its Securities on the

NASDAQ and the PCX. The Company

does not see any particular advantage in

the dual trading of its Securities and

believes that dual listing will fragment

the market for its Securities.

By letter, the Exchange informed the

Company that it has no objection to the

withdrawal of the Company’s Securities

from listing and registration on the PCX.

By reason of Section 12 of the Act and

the rules and regulations thereunder,

the Company shall continue to be

obligated to file reports under Section

13 of the Act with the Commission.

Any interested person may, on or

before May 21, 1998, submit by letter to

the Secretary of the Securities and

Exchange Commission, 450 Fifth Street,

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

bearing upon whether the application

has been made in accordance with the

rules of the Exchange and what terms,

if any, should be imposed by the

Commission for the protection of

investors. The Commission, based on

the information submitted to it, will

issue an order granting the application

after the date mentioned above, unless

the Commission determines to order a

hearing on the matter.

For the Commission, by the Division of

Market Regulation, pursuant to delegated

authority.

Jonathan G. Katz,

Secretary.

[FR Doc. 98–11988 Filed 5–5–98; 8:45 am]

BILLING CODE 8010–01–M

SECURITIES AND EXCHANGE

COMMISSION

[File No. 500–1]

Solucorp Industries, Ltd.; Order of

Suspension of Trading

April 30, 1998

It appears to the Securities and

Exchange Commission that there is a

lack of current and accurate information

concerning the securities of Solucorp

Industries Ltd. (‘‘Solucorp’’) because of

questions regarding the accuracy of

assertions by Solucorp in documents

sent to and statements made to market

makers of the stock of Solucorp, other

broker dealers, and to investors

concerning, among other things: (1) the

negotiation, existence and terms of

contracts entered into by Solucorp

during the period July 1, 1995 through

the present; (2) revenues purportedly

accrued under a license agreement with

Smart International Ltd. and reported in

financial statements for the quarter

ended September 30, 1997 and the sixmonth period ended December 31, 1997,

which were included in a registration

statement and transition report filed

with the Commission in December 1997

and April 1998, respectively; and (3)

revenues projected in press releases on

August 27, 1997, October 24, 1997 and

April 16, 1998.

The Commission is of the opinion that

the public interest and the protection of

investors require a suspension of trading

in the securities of the above listed

company.

Therefore, it is ordered, pursuant to

Section 12(k) of the Securities Exchange

Act of 1934, that trading in the above

listed company is suspended for the

period from 9:30 a.m. EST, May 1, 1998

through 11:59 p.m. EST, on May 14,

1998.

By the Commission.

Jonathan G. Katz,

Secretary.

[FR Doc. 98–12060 Filed 5–1–98; 3:53 pm]

BILLING CODE 8010–01–M