19964 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Westinghouse Electric Company

proprietary information pursuant to 5

U.S.C. 552b(c)(4).

The agenda for the subject meeting

shall be as follows:

Monday, May 11, 1998—8:30 a.m.

until the conclusion of business.

Tuesday, May 12, 1998—8:30 a.m.

until the conclusion of business.

The Subcommittee will continue its

review of the results of the

Westinghouse Test and Analysis

Program supporting the AP600 design

certification. Specifically, the

Subcommittee will review issues

pertaining to the AP600 Reactor Coolant

System, including the resolution of

issues identified in the February 19,

1998 ACRS letter. The purpose of this

meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the

Westinghouse Electric Company, the

NRC staff, their consultants, and other

interested persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, the

scheduling of sessions which are open

to the public, the Chairman’s ruling on

requests for the opportunity to present

oral statements and the time allotted

therefor, can be obtained by contacting

the cognizant ACRS staff engineer, Mr.

Paul A. Boehnert (telephone 301/415–

8065) between 7:30 a.m. and 4:15 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any potential changes to the agenda,

etc., that may have occurred.

Dated: April 16, 1998.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 98–10663 Filed 4–21–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards; Subcommittee Meeting on

Advanced Reactor Designs; Notice of

Meeting

The ACRS Subcommittee on

Advanced Reactor Designs will hold a

meeting on May 13–15, 1998, Room T–

2B3, 11545 Rockville Pike, Rockville,

Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Wednesday, May 13, 1998—8:30 a.m.

until the conclusion of business.

Thursday, May 14, 1998—8:30 a.m.

until the conclusion of business.

Friday, May 15, 1998—8:30 a.m. until

the conclusion of business.

The Subcommittee will continue its

review of the Westinghouse AP600

design. Specifically, the Subcommittee

will review Chapters 3, 6, 9A, 14, 16,

and 17 of the AP600 Standard Safety

Analysis Report, the probabilistic risk

assessment (PRA), regulatory treatment

of non-safety systems (RTNSS), and the

associated NRC staff’s draft Final Safety

Evaluation Report. The purpose of this

meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

Westinghouse Electric Company, their

consultants, and other interested

persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, and

the Chairman’s ruling on requests for

the opportunity to present oral

statements and the time allotted

therefor, can be obtained by contacting

the cognizant ACRS staff engineer, Mr.

Noel F. Dudley (telephone 301/415–

6888) between 7:30 a.m. and 4:15 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any potential changes to the agenda,

etc., that may have occurred.

Dated: April 16, 1998.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 98–10664 Filed 4–21–98; 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.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from March 30,

1998, through April 10, 1998. The last

biweekly notice was published on April

8, 1998 (63 FR 17219).
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19965

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

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 May 22, 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
19966 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Date of amendment request: March

14, 1997.

Description of amendment request:

The proposed amendment would delete

license conditions which have been

satisfied, revise others to delete parts

which are no longer applicable or to

revise references, and make editorial

changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The initial conditions and methodologies

used in the accident analyses remain

unchanged. The proposed changes do not

change or alter the design assumptions for

the systems or components used to mitigate

the consequences of an accident. Therefore,

accident analyses results are not impacted.

The license conditions were one-time

commitments that have been satisfied. There

are no physical changes to the facility, and

all operating procedures, limiting conditions

for operation, limiting safety system settings,

and safety limits are unchanged. Removal of

these license conditions is appropriate and

safe.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Many of the proposed changes delete

references to items that have been completed.

The NRC required these items as a condition

of granting the license. Since they have been

satisfied as intended, deleting them is

administrative.

None of the proposed changes affect the

design or operation of any system, structure,

or component in the plant. The safety

functions of the related structures, systems,

or components are not changed in any

manner, nor is the reliability of any structure,

system, or component reduced by the revised

surveillance or testing requirements. The

changes do not affect the manner by which

the facility is operated and do not change any

facility design feature, structure, system, or

component. No new or different type of

equipment will be installed. Since there is no

change to the facility or operating

procedures, and the safety functions and

reliability of structures, systems or

components are not affected, the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated. The

remaining changes are editorial in nature and

have no impact on plant operation or design.

C. The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes to the Operating

License are generally administrative in

nature and have no impact on the margin of

safety of any Technical Specification. There

is no impact on safety limits or limiting

safety system settings. The changes do not

affect any plant safety parameters or

setpoints. The operating license conditions

have been satisfied, as required. There are no

changes to the conditions themselves.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

Therefore, based on the above evaluation,

Commonwealth Edison has concluded that

these changes do not involve significant

hazards considerations.

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: Byron Public Library District,

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

Illinois 61010.

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. STN 50–454 and STN 50–

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

Ogle County, Illinois

Date of amendment request: October

16, 1997.

Description of amendment request:

The proposed amendment would add an

exemption from 10 CFR 70.24(a) to the

Unit 1 license consistent with the Unit

2 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:

A. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The initial conditions and methodologies

used in the accident analyses remain

unchanged. The proposed change does not

change or alter the design assumptions for

the systems or components used to mitigate

the consequences of an accident. Therefore,

accidents analysis results are not impacted.

There are no physical changes to the

facility, and all operating procedures,

limiting conditions for operation, limiting

safety system settings, and safety limits are

unchanged.

The specific requirements for granting an

exemption from 10 CFR 70.24(a) have been

met. The request is authorized by law, will

not endanger life or property or the common

defense and security, and is in the public

interest.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. 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 affect the

design or operation of any system, structure,

or component in the plant. The safety

functions of structures, systems, or

components are not changed in any manner,

nor is the reliability of any structure, system,

or component reduced by the revised

surveillance or testing requirements. The

change does not affect the manner by which

the facility is operated and does not change

any facility design feature, structure, system,

or component. No new or different type of

equipment will be installed. Since there is no

change to the facility or operating

procedures, and the safety functions and

reliability of structures, systems, or

components are not affected, the proposed

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

C. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change to the Operating

License has no impact on the margin of safety

of any Technical Specification. There is not
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19967

impact on safety limits or limiting safety

system settings. The change does not affect

any plant safety parameters or setpoints.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

Therefore, based on the above evaluation,

Commonwealth Edison has concluded that

the proposed change does not involve

significant hazards considerations.

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: Byron Public Library District,

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

Illinois 61010.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Stuart A.

Richards.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: March

13, 1998, as supplemented March 30,

1998.

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

allow any two auxiliary feedwater

(AFW) flow control valves to be

inoperable concurrently for up to 72

hours, provided the corresponding

redundant flow control valves and a

pump in the other AFW train are

operable.

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:

Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change would only alter the

allowance for specific AFW flow control

valves to be inoperable. It would not affect

any operating limits, any plant operating

conditions, or the physical capability of any

plant equipment. Therefore, it would not

affect the probability of any accident

previously evaluated.

The proposed change would not reduce the

AFW flow capability to the steam generators

during operation under the affected Action

Statement. It would allow more operational

flexibility in plant operation when two AFW

flow control valves in the same train were

concurrently inoperable. The specified AOT

[allowed outage time] of 72 hours would

remain unchanged. Current TS allow

continued operation for 72 hours with one of

the three AFW pumps inoperable, or with

one flow control valve in each train

inoperable (provided the corresponding

redundant flow control valve and a pump in

the other pipe train are operable), but do not

allow continued operation with both valves

in the same train inoperable. The proposed

change would allow any two valves to be

inoperable, with the same provision that the

corresponding redundant flow control valve

and a pump in the other pipe train are

operable.

Since, with the proposed change there

would be no reduction in the ability to

provide AFW flow to either steam generator,

operation of the Facility in accordance with

the proposed changes would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Do the proposed changes create the

possibility of a new or different kind of

accident from any previously evaluated?

The changes do not alter the plant

configuration (no new or different type of

equipment will be installed) or make changes

in the methods governing normal plant

operation. The changes do allow different

sets of AFW flow control valves to be

inoperable, however, these changes retain a

consistent level of AFW capability during

operation under the Action Statement.

Therefore, the changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Therefore, operation of the Facility in

accordance with the proposed TS change

would not create the possibility of a new or

different kind of accident from any

previously evaluated.

Do the proposed changes involve a

significant reduction in a margin of safety?

The proposed change would not reduce the

AFW flow capability to the steam generators

during operation under the affected Action

Statement. It would allow more operational

flexibility in plant operation when two AFW

flow control valves were concurrently

inoperable. The specified AOT of 72 hours

would remain unchanged.

Therefore, operation of the Facility in

accordance with the proposed TS change

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Energy Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: Cynthia A.

Carpenter.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: March

27, 1998 (NRC–98–0033).

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3.5.2,

‘‘ECCS—Shutdown,’’ and TS 3.5.3,

‘‘Suppression Chamber,’’ raising the

minimum water level required in the

condensate storage tank (CST) to

support the core spray system (CSS)

when the suppression pool (the normal

supply for CSS) is unavailable. The

amendment would also eliminate

incorrect information concerning CST

inventory reserved for the high pressure

coolant injection (HPCI) and reactor

core isolation cooling (RCIC) systems.

The associated Bases are also revised.

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 changes will not affect the

performance or reliability of the Condensate

Storage System which could lead to an

accident because the Condensate Storage

Tank (CST) is not involved as an initiator of

any accident previously evaluated. The

proposed change meets the design standards

of the Condensate Storage System by

providing assurance that sufficient water

volume is available for the Core Spray

System. This change also removes [an]

erroneous discussion of water inventory for

HPCI/RCIC Systems while in Operating

[Operational] Conditions 4 and 5. The

removal of information is acceptable since

HPCI/RCIC Systems are not operable in these

modes and will therefore not increase the

probability of an accident. The increase in

volume provides for vortex/air entrainment

avoidance in the Core Spray System and will

not increase consequences. Furthermore, the

elimination of HPCI/RCIC information will

not increase consequences of an accident

previously evaluated because these systems

are not credited for accident mitigation in

Operating [Operational] Conditions 4 and 5.

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 add or

modify any equipment or components related

to the Condensate Storage System and will

therefore not create any new failure modes or

common failure modes. This proposed

change raises the water level within the CST

to ensure sufficient water volume is

maintained and updates the TS by removing

descriptive information with respect to CST

water inventory for HPCI/RCIC Systems
19968 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

while in Operating [Operational] Conditions

4 and 5. The Condensate Storage System will

continue to operate as intended and as

designed. This change will therefore not

create the possibility of a new or different

kind of accident.

3. The change does not involve a

significant reduction in the margin of safety.

The proposed change increases the

required CST water level to provide at least

150,000 gallons of water available for the

Core Spray System while maintaining

adequate submergence of the Core Spray

standpipe for avoiding vortex and air

entrainment. As such, the proposed change

involves no reduction on any margin of

safety. Revision to TS Bases concerning

discussion of reserve volume in CST for HPCI

and RCIC, does not alter the requirement for

Core Spray or Suppression Pool operability

and does not involve a reduction in 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 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: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Project Director: Cynthia A.

Carpenter.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: March

27, 1998 (NRC–98–0034).

Description of amendment request:

The proposed amendment would clarify

a footnote in Technical Specification

(TS) 3.5.1, ‘‘ECCS—Operating,’’ and

3.5.2, ‘‘ECCS—Shutdown,’’ to indicate

that a low pressure coolant injection

system loop may be considered operable

during alignment and operation for

decay heat removal if it is capable of

being manually realigned and is not

otherwise inoperable. The associated

Bases would also be revised.

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 changes involve actions required to

realign the Low Pressure Coolant Injection

(LPCI) system for LPCI injection if LPCI is

required when operating in the Shutdown

Cooling (SDC) mode. The additional actions

described involve resetting isolations and

trips which could occur prior to LPCI

initiation. Resetting these logics does not

initiate any valve operation or pump start;

the LPCI initiation signals and interlocks

remain in control of valve and pump logic.

The equipment interlocks that provide the

isolation signal for the LPCI injection valves

were designed to prevent drain down of the

Reactor Pressure Vessel (RPV) when in SDC.

The injection valve closure is the most

conservative action in response to an RPV

drain event. The current TS acknowledges

that operator action to realign the suction

path is necessary. The proposed change

acknowledges that operator action to reset

injection valve logic and pump trips is

necessary. The time required to realign LPCI

is not significantly different than the existing

actions to realign the suction path.

No changes in either system design or

operating strategies will be made as a result

of these changes, thus no opportunity exists

to increase the probability or consequences of

a previously analyzed accident. Therefore,

the proposed change does not involve a

significant increase in the probability or

consequences of a previously evaluated

accident.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The manual realignment of the LPCI

system from SDC following an isolation

signal does not affect the accident analysis

described in Chapter 15 of the UFSAR

[updated final safety analysis report]. No new

limiting single failure has been identified as

a result of the proposed changes. The

possibility of a new or different kind of

accident from those previously analyzed will

not be created by the change to the TS

footnote or Bases, because the proposed

change merely clarifies the actions necessary

to realign the LPCI system. The time required

to realign the system is not significantly

different than the time necessary to realign

the suction path. Therefore, no new or

different types of failures or accident

initiators are introduced by the proposed

changes.

3. The change does not involve a

significant reduction in the margin of safety.

The proposed change described above

affects the plant’s ability to enter Operational

Conditions 3, 4, and 5, and to achieve and

maintain COLD SHUTDOWN conditions

when shutting down the plant. The proposed

change in combination with existing

restrictions within the TS provide assurance

that there is no credible mechanism to inhibit

running the LPCI system. The minor

additional operator action required to realign

LPCI from SDC requires minimum time and

effort considering controls for each division

are located on their respective control panel.

As a result of this change, there will be no

changes in either system design or operating

strategies because the proposed changes

merely clarify existing TS requirements and

actions necessary to meet TS requirements.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Project Director: Cynthia A.

Carpenter.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: April 2,

1998 (NRC–98–0057).

Description of amendment request:

The proposed amendment would permit

entering Operational Conditions 1 and 2

prior to completion of Surveillance

Requirements for the primary

containment hydrogen and oxygen

monitors in order to establish the

conditions necessary (inerted

containment) to properly perform the

calibrations. The amendment would

also increase the frequency of the

calibration for the oxygen monitors from

every 18 months to quarterly in

accordance with vendor

recommendations and correct the

nomenclature for the hydrogen and

oxygen monitors in tables 3.3.7.5–1 and

4.3.7.5–1.

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. The changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change will permit delaying

the performance of calibrations of the

hydrogen and oxygen monitors until after the

containment is inerted following a plant

startup. The proposed change will also

increase the calibration frequency for the

oxygen monitors from once per 18 months to

once per quarter, and change the

nomenclature for the hydrogen and oxygen

monitors.

The primary containment hydrogen and

oxygen monitors are passive instruments that

provide indication to control room operators

of hydrogen and oxygen concentration in the

primary containment. Because they perform

only a passive monitoring function, the
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19969

hydrogen and oxygen monitors are not

associated with the initiation of any

previously evaluated accident. The

indication provided by the monitors is used

by the control room operators to ensure

oxygen concentration remains below limits

and to make decisions regarding the use of

the Combustible Gas Control System, if

necessary. The allowance to permit entry into

applicable operational conditions before

calibration ensures that the conditions

(nitrogen environment) are appropriate for

accurate calibration of the instrument.

Delaying the calibration does not cause the

instrument to cease to function. Calibrations

verify and adjust, as necessary, the accuracy

of the instrument to compensate for drift that

may occur since the last calibration. Thus,

even with a delayed calibration, the

instruments still would provide valuable

information to the operators. Consequently,

this change will not involve a significant

increase in the consequences of a previously

evaluated accident because the monitors will

still function and provide meaningful

information until the calibration is

completed.

The change to reduce the interval for

calibration of the oxygen monitors from once

per 18 months to once per quarter provides

increased assurance of monitor accuracy and

is consistent with the manufacturer’s

recommendations. Therefore, because this

instrument is not associated with the

initiation of an accident and the change

improves the functionality of the instrument,

the probability and consequences of

previously evaluated accidents are not

significantly affected.

The change in nomenclature is editorial,

and, as such does not affect the probability

or consequences of a previously evaluated

accident.

2. The changes do not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

As discussed above, the hydrogen and

oxygen monitors are passive, indication-only

instruments which provide information to

control room operators. The proposed

changes do not introduce a new mode of

operation or involve a physical modification

to the plant. Therefore, the proposed changes

do not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. The changes do not involve a significant

reduction in the margin of safety.

The proposed changes involve the

containment hydrogen and oxygen monitors

which do not affect any parameters or

assumptions used in the calculation of any

safety margin with regard to Technical

Specification Safety Limits, Limiting Safety

System Settings, Limiting Control Settings or

Limiting Conditions for Operation, or other

previously defined margins for any structure,

system, or component. Therefore, the

proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Project Director: Cynthia A.

Carpenter.

Duquesne Light Company, et al., Docket

Nos. 50–334 and 50–412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

(BVPS–1 and BVPS–2), Shippingport,

Pennsylvania

Date of amendment request: March

16, 1998.

Description of amendment request:

The proposed amendment would revise

technical specification (TS) Table 4.3–1

to add footnote 6 to the channel

calibration requirement for all

instrument channels that are provided

with an input from neutron flux

detectors. Footnote 6 provides that

neutron detectors may be excluded from

channel calibrations. Additional

changes are proposed for BVPS–1 to

provide consistency between BVPS–1

and BVPS–2. These additional changes

would add channel calibration

requirements to BVPS–1 TS Table 4.3–

1 items 2.b. (Power Range, Neutron

Flux, Low Setpoint), 5. (Intermediate

Range, Neutron Flux), 6. (Source Range,

Neutron Flux (Below P–10), and 23.

(Reactor Trip System Interlocks P–6, P–

8, P–9, and P–10). Furthermore, changes

would be made to correct page numbers

in the BVPS–2 Index and to add

corresponding changes to the bases for

both units.

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?

Overall protection system performance will

remain within the bounds of the previously

performed accident analyses since no

hardware changes are proposed. The

protection systems will continue to function

in a manner consistent with the plant design

basis. The proposed changes will not affect

any of the analysis assumptions for any of the

accidents previously evaluated. The

proposed changes will not affect the

probability of any event initiators nor will

the proposed changes affect the ability of any

safety-related equipment to perform its

intended function. There will be no

degradation in the performance of nor an

increase in the challenges imposed on safetyrelated equipment assumed to function

during an accident. There will be no change

to normal plant operating parameters or

accident mitigation capabilities. Therefore,

the proposed changes do 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?

There are no hardware changes associated

with this license amendment nor are there

any changes in the method by which any

safety-related plant system performs its safety

function. The normal manner of plant

operation is unchanged.

No new accident scenarios, transient

precursors, failure mechanisms, or limiting

single failures are introduced as a result of

these changes. There will be no adverse effect

or challenges imposed on any safety-related

system as a result of these changes.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed change does not involve a

significant reduction in a margin of safety.

The proposed changes do not affect the

acceptance criteria for any analyzed event

nor is there a change to any Safety Analysis

Limit (SAL). Maintaining the SAL preserves

the margin of safety.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems necessary to

assure the accomplishment of protection

functions. Therefore, the proposed changes

do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B.F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Robert A. Capra.

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 amendment request: March

17, 1998.

Description of amendment request:

The proposed amendment would revise

Action 34 of technical specification (TS)
19970 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Table 3.3–3, ‘‘Engineered Safety Feature

Actuation System Instrumentation.’’

Action 34 applies to Functional Units

6.b., ‘‘Grid Degraded Voltage (4.16 kV

Bus),’’ and 6.c. ‘‘Grid Degraded Voltage

(480 v Bus).’’ The proposed revision

would require that with one channel

inoperable, the inoperable channel be

placed in the tripped condition within

one hour; otherwise, the applicable

action statement(s) for the associated

emergency diesel generator made

inoperable by the degraded voltage start

instrumentation be entered

immediately. The proposed revision

would also require that with two

channels inoperable, at least one of the

two channels be restored to operable

status and the other channel be placed

in the tripped condition within one

hour; otherwise, the associated

emergency diesel generator shall be

declared inoperable and its applicable

action statement(s) shall be entered. In

addition, corresponding changes would

be made to the bases for TS 3/4.3.2 and

the BVPS–2 Index would be revised to

reflect changed page numbers.

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?

Engineered Safety Feature Actuation

System (ESFAS) will continue to function in

a manner consistent with the plant design

basis. The proposed change will not affect

any of the analysis assumptions for any of the

accidents previously evaluated. The

proposed changes will not affect the

probability of any event initiators. There will

be no change to normal plant operating

parameters. The emergency bus degraded

voltage protection system is utilized for

accident mitigation and is not considered to

be the source of accidents previously

evaluated.

Implementation of the proposed changes

will now provide viable corrective actions

which do not significantly increase the

probability of failure of safety related

equipment to perform its intended function.

The proposed Action 34 permits a one hour

time frame before the affected diesel

generator(s) is required to be declared

inoperable. This one hour period allows for

repairs of most failures and takes into

account the low probability of an event

which would require the degraded voltage

protection system to function. If adequate

protection is not restored within this one

hour period, the diesel generator(s) allowable

outage time is invoked. The diesel

generator(s) allowable outage time has been

previously evaluated and determined to be

an acceptable period of time during which

plant operation may continue without an

emergency backup power source. The loss of

emergency bus degraded voltage protection is

similar to the loss of the ability of an

emergency diesel generator to provide

electrical power to the safety related loads on

the emergency buses. In both situations, a

loss of offsite power, due to a total loss or a

degraded condition, will result in the safety

related loads not being capable of mitigating

a design basis accident. The proposed

changes to the Index page are administrative

in nature and do not affect plant safety.

Therefore, the proposed changes do not

result in a significant increase in 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 emergency bus degraded voltage

protection system is utilized for accident

mitigation. The proposed changes will now

provide viable corrective actions which do

not result in a change in the manner in which

the emergency bus loads are protected from

a degraded voltage condition. These changes

do not alter the function of the degraded

voltage protection system. The proposed

changes will continue to require that at least

one of the two redundant 4160 volt or 480

volt emergency buses is protected from a

degraded voltage condition assuming a single

active failure of the opposite emergency bus

degraded voltage protection system. This

action will ensure that at least one train of

engineered safety feature (ESF) equipment is

not damaged due to a sustained bus

undervoltage condition. The proposed

addition of the requirement to enter the

action statement for the inoperable diesel

generator, if the one hour requirements of

Action 34 cannot be met, will ensure that

adequate compensatory actions to assure

plant safety are taken. These requirements

include the demonstration of the operability

of the A.C. offsite sources by performing a

specific surveillance within one hour and at

least once per eight hours thereafter. If both

diesel generators are inoperable, at least one

diesel generator must be restored to operable

status within two hours or the plant must be

placed in cold shutdown within the

following 36 hours.

No new accident scenarios, failure

mechanisms, or limiting single failures are

introduced as a result of these changes. There

will be no adverse effect or challenges

imposed on any safety-related system as a

result of these changes.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The margin of safety is not significantly

reduced because the A.C. electrical power

sources will continue to provide sufficient

capability, redundancy, and reliability to

ensure availability of necessary power to ESF

systems. The ESF systems will continue to

function, as assumed in the safety analyses,

to ensure that fuel, reactor coolant system

and containment design limits are not

exceeded. The proposed revisions to Action

34 will continue to require that at least one

of the two redundant 4160 volt or 480 volt

emergency buses is protected from a

degraded voltage condition assuming a single

active failure of the opposite emergency bus

degraded voltage protection system. This

action will ensure that at least one train of

ESF equipment is not damaged due to a

sustained bus undervoltage condition. The

emergency loads, which are powered from

that train of emergency buses, will continue

to be available to perform their safety related

functions. If the one hour requirements of

Action 34 cannot be met, the affected

emergency diesel generator will be declared

inoperable. This will ensure that adequate

compensatory actions to ensure plant safety

are taken. The loss of emergency bus

protection from a degraded voltage condition

is similar to the loss of the ability of an

emergency diesel generator to provide

electrical power to the safety related loads on

the emergency buses. In both situations, a

loss of the offsite power, due to a total loss

or a degraded condition, will result in the

safety related loads not being capable of

mitigating a design basis accident.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B.F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Robert A. Capra.

GPU Nuclear Corporation, et al., Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County,

Pennsylvania

Date of amendment request: March

23, 1998.

Description of amendment request:

The proposed amendment modifies

Section 3.1.2 of the Technical

Specifications (TS) to incorporate new

pressure/temperature limits regarding

reactor vessel pressurization heatup,

cooldown, and inservice leak and

hydrostatic leak test limitations in

accordance with 10 CFR 50, Appendix

G. These new limits would be

applicable through the period of 17.7

effective full power years (EFPY) of

operation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19971

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or consequences of

an accident previously evaluated. The design

basis event related to this change is

nonductile failure of the reactor coolant

pressure boundary. The updated pressure/

temperature limits have been established in

accordance with the requirements of 10 CFR

50, Appendix G. Revision of these curves for

an applicability period of 17.7 EFPY is based

on maintaining the required design margin.

Operation of the facility in accordance with

the proposed amendment provides assurance

of protection against nonductile failure of the

reactor coolant pressure boundary for

operation through 17.7 EFPY. Therefore,

operation in accordance with the proposed

amendment does not involve a significant

increase in the probability of occurrence or

consequences of an accident previously

evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any previously

evaluated. The design basis event related to

the change is nonductile failure of the reactor

coolant boundary. The proposed amendment

provides assurance of protection against

nonductile failure of the reactor coolant

boundary for operation through 17.7 EFPY

and is unrelated to the possibility of creating

a new or different kind of accident.

3. Operation of the facility in accordance

with the proposed amendment would not

involve any reduction in a margin of safety

since the design methodology has maintained

the existing margins.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Cecil O.

Thomas.

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 revise the

Limiting Conditions for Operation in the

event that one 250 V DC electrical

power subsystem is inoperable.

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 DC electrical power sources are used

to support mitigation of the consequences of

an accident; however, they are not

considered the initiator of any previously

analyzed accident. The proposed change

merely provides direction to the operator to

declare equipment associated with a 250 V

DC electrical power subsystem inoperable if

the subsystem becomes inoperable. This

provides assurance that all affected features

are immediately recognized as incapable of

performing their safety functions, and

requires immediate actions equivalent to

those determined appropriate in the

Technical Specifications for the affected

features. 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 introduce a

new mode of plant operation and does not

involve physical modification to the plant.

Therefore, 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?

This change does not involve a significant

reduction in a margin of safety, since the

proposed change results in establishing the

level of safety for the loss of a 250 V DC

electrical power subsystem equivalent to the

level of safety that exists in the Technical

Specifications for components and systems

that are supplied by the 250 V DC electrical

power system.

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.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: March 2,

1998.

Description of amendment request:

The proposed changes would revise the

frequency for the performance of

specific surveillances associated with

the emergency diesel generators (EDGs)

and delete the requirements contained

in the current Technical Specifications

for accelerated testing whenever the

number of valid test failures associated

with the EDGs is met or exceeded. In

addition, the special requirements for

reporting valid or invalid EDG failures

would 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:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not affect

accident initiators or precursors and do not

alter the design assumptions affecting the

ability of the EDGs to mitigate the

consequences of an accident.

Industry experience has indicated that

excessive testing requirements have proven

to be a contributor to increased EDG

unavailability and equipment degradation.

Removing inappropriate testing requirements

increases EDG reliability and enhances the

ability of EDGs to mitigate the consequences

of an accident. Implementing the

maintenance rule in accordance with 10 CFR

50.65, Regulatory Guide 1.160, and NUMARC

93–01 for the EDGs provides additional

assurance that high EDG performance and

availability will be maintained.

Deleting the special reporting requirements

from the Technical Specifications is an

administrative change that does not affect the

ability of the EDGs to perform their specified

safety function. North Atlantic will continue

to notify the NRC of significant EDG failures

in accordance with the provisions of 10 CFR

50.72 and 50.73.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed changes do not alter the

ability of the EDGs 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 have no impact on component or

system interactions, or the plant design basis.

Instrumentation setpoints, starting,
19972 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

sequencing and loading functions associated

with the EDGs are not affected by the

proposed changes. Furthermore, combining

the implementation of the maintenance rule

program with the proposed amendment will

enhance both the availability and the

performance of the EDGs.

Therefore, 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.

There is no impact 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 do not affect the EDG’s

ability to ensure that sufficient power is

available to supply the safety related

equipment required for: 1) the safe shutdown

of the facility, and 2) the mitigation and

control of accident conditions within the

facility. In addition, the proposed changes do

not affect the EDG’s ability to ensure that: 1)

the facility can be maintained in a shutdown

or refueling condition for extended periods of

time, and 2) sufficient instrumentation and

control capability is available for monitoring

and maintaining the unit status.

EDG reliability and availability are

expected to be improved by the proposed

changes. Eliminating excessive testing

requirements can improve safety by reducing

challenges to plant systems and reducing

equipment wear and degradation. While the

proposed changes affect surveillance

intervals there are no changes to the methods

used to perform the surveillances. The

surveillances will continue to demonstrate

the ability of the EDGs to perform their

intended function of providing electrical

power to the emergency safety systems

needed to mitigate design basis transients

and accidents. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 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,

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

NRC Project Director: Cecil O.

Thomas.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: March 5,

1998.

Description of amendment request:

The proposed changes would revise the

Seabrook Station Radiological Effluent

Technical Specifications (TS) and

Administrative Controls section of the

Technical Specifications, as authorized

by NRC Generic Letter (GL) 89–01,

‘‘Implementation Of Programmatic

Controls For Radiological Effluent

Technical Specifications (RETS) In The

Administrative Controls Section Of The

Technical Specifications And The

Relocation Of Procedural Details of

RETS To The Offsite Dose Calculation

Manual Or To The Process Control

Program.’’ The proposed amendment

would incorporate programmatic

controls in the TSs for radioactive

effluents and for environmental

monitoring conforming to the applicable

regulatory requirements and would

relocate the existing procedural details

of the current RETS to the Offsite Dose

Calculation Manual (ODCM). Procedural

details associated with solid radioactive

wastes would be relocated to the

Process Control Program (PCP).

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.

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

structures, systems, or components (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 and

procedural details relative to radiological

effluents.

Incorporation of programmatic controls for

RETS in TSs will assure that the applicable

regulatory requirements pertaining to the

control of radioactive effluents will continue

to be maintained. 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.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed changes do not alter the

design assumptions, conditions,

configuration of the facility or the manner in

which the plant is operated. 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 relative to radiological

effluents. Therefore, 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.

There is no impact 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

relative to radiological effluents. A

comparable level of administrative control

will continue to be applied to those design

conditions and associated surveillances being

relocated to the ODCM or PCP. Therefore, the

proposed changes do not involve a

significant reduction in 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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Cecil O.

Thomas.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: March

23, 1998.

Description of amendment request:

The proposed change would revise the

Seabrook Station Technical

Specifications (TSs) to add a new TS

3.0.5 that would provide an exception to

TSs 3.0.1 and 3.0.2 to allow the

performance of required testing to

demonstrate the operability of the

equipment being returned to service or

the operability of other equipment.

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.

The design basis accidents are not affected

by the proposed administrative changes.
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19973

Specification 3.0.5 provides the

administrative controls to ensure the time the

equipment is returned to service in conflict

with the requirements of the ACTIONS is

limited to the time absolutely necessary to

perform the allowed required testing.

Specification 3.0.5 was incorporated in

NUREG–1431, ‘‘Standard Technical

Specifications—Westinghouse Plants,’’ (as

modified by approved Technical

Specification Task Force (TSTF) generic

change Traveler TSTF–165), to address these,

and other similar situations, that conflict

with the requirements with the ACTIONS

when equipment is returned to service.

Specification 3.0.5 does not provide time to

perform other preventative or corrective

maintenance.

Inclusion of Specification 3.0.5 into the

Seabrook Station Technical Specifications

will provide operational flexibility with the

restrictive compliance requirements of the

other Applicability Specifications (3.0.1 and

3.0.2) and allow the performance of postmaintenance/surveillance activities to

facilitate returning equipment to service or to

allow other equipment to be tested.

Therefore, inclusion of Specification 3.0.5

into the Seabrook Station Technical

Specifications enhances plant safety by

minimizing the potential for plant trip and/

or transients. A qualitative risk assessment

concerning returning components to service

for post-maintenance testing was performed

and concluded that the configurations

allowed by Specification 3.0.5 have a

negligible effect on the Seabrook Station risk

profile. The components involved will have

either completed calibration or maintenance,

and can reasonably be expected to be able to

perform their required safety function when

returned to service for testing purposes.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed changes do not introduce

new features or modify plant structures,

systems and components or procedures that

could possibly affect station operations under

normal or abnormal conditions, thus, the

potential for an unanalyzed accident is not

created. The proposed administrative

changes have no adverse affect on the safety

limits or design basis accidents. Therefore,

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.

There are no changes being made to the

Technical Specification safety limits or safety

system settings that would adversely affect

plant safety. The changes do not affect the

operation of structures, systems or

components (SSCs) nor do they introduce

administrative changes to plant procedures

that could affect operator response during

normal, abnormal or emergency situations.

Inclusion of Specification 3.0.5 into the

Seabrook Station Technical Specifications

enhances plant safety by minimizing the

potential for plant trip and/or transients by

allowing equipment to be returned to service.

A qualitative risk assessment concerning the

return of components to service for postmaintenance testing was performed and

concluded that the configurations allowed by

Specification 3.0.5 have a negligible effect on

the Seabrook Station risk profile. The

components involved will have either

completed calibration or maintenance, and

can reasonably be expected to be able to

perform their required safety function when

returned to service for testing purposes.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 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,

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

NRC Project Director: Cecil O.

Thomas.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: March

27, 1998.

Description of amendment request:

The proposed change would revise

Technical Specification (TS) 3.7.6,

‘‘Control Room Emergency Makeup Air

and Filtration (CREMAFS).’’ The

proposed change would modify the

existing required action when both

trains of CREMAFS are inoperable in

Modes 5 and 6 by eliminating the

restriction of suspending positive

reactivity changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes have no impact on

the probability or consequences of an

accident previously evaluated in the UFSAR.

The control room ventilation systems are

support systems which have a role in the

detection and mitigation of accidents but do

not contribute to the initiation of any

accident previously evaluated. The removal

of the positive reactivity addition restriction

in Mode 5 and 6 has no impact on the course

of any accidents previously evaluated. There

are no presently evaluated positive reactivity

or boron dilution accidents that credit the

CREMAFS to mitigate its consequences or

provide radiological protection. The positive

reactivity restriction is overly restrictive in

that it does not allow cooldown below 200°

F when Mode 5 is entered as a result of both

trains of CREMAFS being inoperable nor

does it allow Reactor Coolant System

temperature to vary.

The restriction is also redundant to

Technical Specification 3.1.1.2 ‘‘Reactivity

Control Systems Shutdown Margin-Tavg less

than or equal to 200° F’’ in Mode 5 and

Technical Specification 3.9.1 ‘‘Refueling

Operations Boron Concentration’’ in Mode 6.

Technical Specification 3.1.1.2 action, with

shutdown margin less than the limit

specified in the Core Operating Limits Report

or with the Reactor Coolant System boron

concentration less than 2000 ppm boron,

requires immediate and continued boration

until the restoration of the required

shutdown margin or boron concentration.

Similarly, Technical Specification 3.9.1

actions require suspension of core alterations

or positive reactivity changes in addition to

immediate and continued boration until the

restoration of the required shutdown margin

(Keff) or boron concentration while in Mode

6. Sufficient shutdown margin ensures that

(1) the reactor can be made subcritical from

all operating conditions, (2) the reactivity

transients associated with the postulated

accident conditions are controllable within

acceptable limits and (3) the reactor will be

maintained sufficiently subcritical to

preclude inadvertent criticality in the

shutdown condition. The above referenced

reactivity control system specifications

provide the necessary protection for

postulated reactivity addition accident

conditions. Therefore, modifying the

Technical Specification action that requires

the suspension of positive reactivity changes

and core alterations with both trains of the

CREMAFS inoperable does not involve a

significant increase in the consequences of an

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed change that removes the

positive reactivity addition restriction in

Mode 5 and 6 does not create the possibility

of a new accident nor does it create the

possibility of a different kind of accident

previously evaluated. There are no presently

evaluated positive reactivity or boron

dilution accidents that credit the CREMAFS

to mitigate its consequences or provide

radiological protection. The addition of

positive reactivity during the above described

situation is overly restrictive and furthermore

redundant to Technical Specification 3.1.1.2

‘‘Reactivity Control Systems Shutdown

Margin-Tavg less than or equal to 200° F’’ in

Mode 5 and Technical Specification 3.9.1

‘‘Refueling Operations Boron Concentration’’

in Mode 6. The above referenced reactivity

control system specifications provide the

necessary protection for postulated reactivity

addition accident conditions. Therefore,

modifying the Technical Specification action

that requires the suspension of positive
19974 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

reactivity changes and core alterations with

both trains of the CREMAFS inoperable does

not create the possibility of a new or different

accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in the margin of

safety.

The changes being proposed do not revise

equipment design or operation nor do they

make changes to Technical Specification

required safety limits or safety system

settings. In addition, they do not alter the

environmental conditions which are to be

maintained in the control room during

normal operation and following an accident

and they do not revise the accident analyses.

Therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

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,

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

NRC Project Director: Cecil O.

Thomas.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 3,

1998.

Description of amendment request:

The proposed change would revise the

Seabrook Station Technical

Specifications (TSs) with administrative

changes to support phased

implementation of 24-month fuel cycle

surveillance interval extensions.

Specifically, the proposed change

would: (1) provide wording changes in

the Bases Section of TS 4.0.2 necessary

to support 24-month surveillance

interval extensions, (2) revise TS 4.0.5.b

to provide revised terminology for

inservice inspection and testing

activities and their associated

frequencies, (3) revise TS Table 1.1 to

clarify current and future refueling

intervals and their associated

surveillance requirements and

frequencies, and (4) delete the ‘‘during

shutdown’’ restriction from the

performance requirements of certain

surveillance requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The design basis accidents are not affected

by the proposed editorial and administrative

changes. The proposed changes do not

change the level of programmatic controls or

the procedural details currently in place.

Furthermore, these changes have no adverse

affect to the safe operation of the station.

Performance of certain maintenance and

testing activities during conditions or modes

other than shutdown will be evaluated by

North Atlantic to ensure proper regard to

their effect on safe operation of the plant is

given prior to conduct of a particular

surveillance, or portion thereof. Therefore,

the proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed changes do not introduce

new features or modify plant structures,

systems and components or procedures that

could possibly affect station operations under

normal or abnormal conditions, thus, the

potential for an unanalyzed accident is not

created. Performance of maintenance and

testing activities on-line, as well as

shutdown, are controlled by North Atlantic’s

procedures and policies to perform reviews

and assessments of these activities to

determine the affect on safe operation of the

facility. The proposed editorial and

administrative changes have no adverse

affect on the safety limits or design basis

accidents. Therefore, 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.

There are no changes being made to the

Technical Specification safety limits or safety

system settings that would adversely affect

plant safety. The changes do not affect the

operation of structures, systems or

components nor do they introduce

administrative changes to plant procedures

that could affect operator response during

normal, abnormal or emergency situations.

Performance of certain maintenance and

testing activities during conditions or modes

other than shutdown will be evaluated by

North Atlantic to ensure proper regard to

their effect on safe operation of the plant is

given prior to conduct of a particular

surveillance, or portion thereof. Therefore,

the proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 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,

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

NRC Project Director: Cecil O.

Thomas.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: April 1,

1998.

Description of amendment request:

The proposed revision to the Millstone

Unit 3 licensing basis would add a new

sump pump subsystem to address

groundwater inleakage through the

containment basemat.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

revision in accordance with 10 CFR 50.92

and has concluded that the revision does not

involve a significant hazards consideration

(SHC). The bases for this conclusion is that

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

satisfied. The proposed revision does not

involve an SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The current FSAR [Final Safety Analysis

Report] credits the waterproof membrane for

assuring that groundwater inleakage is not

significant and would have no impact on

safety related structures and components.

However, degradation of the waterproof

membrane has been detected, and it is now

concluded that groundwater inleakage can be

significant in that it could affect the

operability of the RSS [recirculation spray

system] pumps. The original plant design

had only nonsafety-related RSS sump pumps

available for pumping the groundwater from

the RSS sumps. These pumps are not

powered from the emergency busses and

would not be accessible during a design basis

LOCA [loss-of-coolant accident].

Thus, it is assumed that they would not be

available to mitigate a design basis accident.

Two independent safety-related air-driven

sump pumps have been installed to eliminate

the potential for groundwater inleakage that

would affect the RSS pumps.

Air-driven sump pumps have been

installed with the air supply line routed to

a connection outside the ESF [engineered

safety features] building. This allows the

installation of an air compressor in an area

that is accessible during a design basis

accident such as a LOCA. Two air

compressors have been staged in designated

locations, and will be maintained and
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19975

periodically tested to ensure their

availability. Periodic testing of the sump

pumps will also be performed. The

surveillance requirements have been

incorporated into the Technical

Requirements Manual.

EOP [Emergency Operating Procedure] 35–

ES1.3 has been modified to add a step to

install the compressors and start the sump

pumps. It is estimated that these sump

pumps would be needed approximately ten

hours after a design basis accident. Thus,

there is sufficient time for the operators to

perform this action. Since sufficient time is

available, the action has been incorporated

into procedures and the environmental

conditions allow access to the area, it is

concluded that credit for operator action can

be taken.

Thus, the new system is single failure

proof and meets the requirements of

Standard Review Plan 3.4.1 which stated the

following:

‘‘If safety-related structures are protected

from below-grade groundwater seepage by

means of a permanent dewatering system,

then the system should be designed as a

safety-related system and meet the single

failure proof criterion.’’

This provides assurance that the RSS

pumps and other safety-related structures

and components will perform the required

safety function as assumed in the accident

analysis.

The current nonsafety-related RSS sump

pump system will continue to provide

protection from groundwater inleakage

during normal operation. Thus, there is no

impact on the probability of occurrence of a

transient because of equipment or structural

failure due to groundwater inleakage. In

addition, the new safety-related RSS sump

pump system provides additional assurance

that groundwater inleakage would not affect

structures or equipment during an extended

loss of offsite power or a design basis

accident. Thus, it is concluded that there is

no impact on the probability of occurrence of

any previously evaluated accident.

The change results in the use of the new

air-driven sump pumps to remove

groundwater in-leakage from the RSS

cubicles. To preclude the possibility for

radiological contamination of the

groundwater, all sources of liquid

radiological contamination to the sumps have

been eliminated. The RSS cubicle floor

drains leading to Sumps 7A/7B have been

plugged. Drains from equipment determined

not to be a potential source of radiological

contamination continue to drain to Sumps

7A/7B (sources include CCP [component

cooling water] and Service Water relief

valves) and are covered with splash guards

to prevent the entrance of contaminated

spray. The Hydrogen Recombiner area floor

drains and the drain from the PASS [post

accident sampling system] sample sink, all of

which are nonsafety-related, have been

isolated from the indirect waste receptor

which drains to Sump 7B. Sumps 7A and 7B

have been cleaned and the existing

nonsafety-related sump pumps replaced to

remove any existing residual contamination.

The nonsafety-related pumps (3DAS–P8A/B)

discharge to ESF Building sump 3DAS–

SUMP 10. To preclude any potential

siphoning from the potentially contaminated

Sump 10 back to Sumps 7A/7B, the lines of

the existing nonsafety-related pumps have

been shortened to discharge above the water

level in Sump 10.

The walls of Sumps 7A/7B have been

extended to protect from a Limited Passive

Failure and Pipe Break in the RSS cubicles.

The expected flooding height is 6.6 inches

[ ]. The sump cubicle height was extended

to 3 ft. above the cubicle floor, well above

this height. The sumps are covered with a

vented hood to protect from pipe break spray

and miscellaneous overhead leaks to further

assure the sumps remain isolated from

potentially contaminated RSS system fluids.

The existing SLCRS [supplementary leak

collection and release system] boundary has

been extended to the isolation valves located

outside of the ESF building. Additionally,

when the sump level is reduced while using

the air driven pump, the pumps are designed

to prevent air from being discharged through

the pump discharge outside of the ESF

building.

Thus, use of the new sump pumps would

not affect the offsite doses following a design

basis accident.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The current nonsafety-related RSS sump

pump system will continue to provide

protection from groundwater inleakage

during normal operation. This will continue

to provide assurance there is no potential for

a transient because of equipment or

structural failure due to groundwater

inleakage. In addition, the new safety-related

RSS sump pump system provides additional

assurance that groundwater inleakage would

not affect structures or equipment during an

extended loss of offsite power or a design

basis accident.

Therefore, the proposed revision does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The current FSAR credits the waterproof

membrane for assuring that groundwater

inleakage is not significant and would have

no impact on safety related structures and

components. However, degradation of the

waterproof membrane has been detected and

it is now concluded that groundwater

inleakage can be significant in that it could

affect the operability of the RSS pumps.

Original design had only nonsafety-related

RSS sump pumps available for pumping the

groundwater from the RSS sumps. These

pumps are not powered from the emergency

busses and would not be accessible during a

design basis LOCA. Thus, it is assumed that

they would not be available to mitigate a

design basis accident. Two independent

safety-related air-driven sump pumps have

been installed to eliminate the potential for

groundwater inleakage that would affect the

RSS pumps. The new system is single failure

proof and meets the requirements of

Standard Review Plan 3.4.1.

Use of the new system requires operator

action to install pre-staged air compressors to

provide power for the new air-driven sump

pumps. It is estimated that these sump

pumps would be needed approximately ten

hours after a design basis accident. Thus,

there is sufficient time for the operators to

perform this action. Since sufficient time is

available, the action has been incorporated

into procedures and the environmental

conditions allow access to the area, it is

concluded that credit for operator action can

be taken.

With credit for the new single failure proof

air-driven sump pumps and operator action

to install pre-staged compressors to provide

power for the pumps, the new subsystem

provides the required assurance that the RSS

pumps will not be affected by groundwater

inleakage. Thus, it is concluded that the RSS

pumps would be operable for long term

accident mitigation and there is no impact on

the margin of safety as defined in the basis

of the Emergency Core Cooling Technical

Specifications or any other Technical

Specification.

Therefore, the proposed revision does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the

information provided, it is determined

that the proposed revision does not

involve an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

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:

December 23, 1997.

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 Technical Specification (TS) 3/
19976 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

4.7.11, Table 3.7–1, ‘‘Maximum

Allowable Power Range Neutron Flux

High Setpoint With Inoperable Steam

Line Safety Valves.’’ The power range

(PR) neutron flux high setpoints would

be changed based on revised

calculational methodologies for 1, 2, or

3 inoperable MSSVs per steam generator

(SG). The proposed TS change would

lower the PR neutron flux high setpoints

when 2 or 3 MSSVs are inoperable per

loop such that the maximum power

level allowed would be within the heat

removing capability of the remaining

operable MSSVs. Although the method

for calculating the maximum power

level allowed when one MSSV per loop

is inoperable has been revised, the

results have not and the limit remains

the same. The associated Bases would

also be revised.

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 reduction in the power range (PR)

neutron flux high setpoint Technical

Specification (TS) values does not initiate an

accident. Technician adjustments to lower

the PR neutron flux high setpoints could

cause a reactor trip (RT). However, this

action is already a TS requirement. Thus,

reducing the TS setpoint values from their

current values will not change the

requirement for a technician to adjust the

setpoints downward when main steam safety

valves (MSSVs) become inoperable, and

therefore, will not increase the probability of

a RT.

The reduction of the setpoints assures that

the consequences of an accident when the

MSSVs are inoperable are not affected by

assuring that the MSSVs will continue to

prevent overpressure of the main steam leads

and steam generators (SGs) and remove

adequate heat for the reactor coolant system.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Reduction of the PR neutron flux high

setpoints does not change the method by

which any safety-related system performs the

function.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

With the reduction in the PR neutron flux

high setpoints for inoperable MSSVs, the

MSSVs will still prevent SG pressure from

exceeding 110 percent of SG design pressure

in accordance with the ASME code. The

change is conservative. The conclusions for

the Final Safety Analysis Report Update

accident analyses are unaffected by the

change, remain valid, and provide margin.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment 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 and Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Project Director: William H.

Bateman.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: February

6, 1998.

Description of amendment request:

The proposed changes would revise the

Reactor Protection System (RPS) Normal

Supply Electrical Protection Assembly

(EPA) Undervoltage Trip setpoint to

reflect a reanalysis of the most limiting

applied load minimum voltage

requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed setpoint change evaluated in

Section III does not involve any physical

changes to the plant, does not alter the way

these systems function, and will not degrade

the performance of the plant safety systems.

The proposed instrument setpoint changes

ensures that plant safety limits are not

exceeded for the most limiting voltage

requirements. The type of testing and the

corrective actions required if the subject

surveillances fail remains the same. The

proposed changes do not adversely affect the

reliability of these systems or affect the

ability of the systems to meet their design

objectives. A historical review of surveillance

test results supports these conclusions.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed setpoint change evaluated in

Section III does not modify the design or

operation of the plant, therefore, no new

failure modes are introduced. The proposed

instrument setpoint change ensures that

plant safety limits are not exceeded for the

most limiting voltage requirements. No

changes are proposed to the type and method

of testing performed. A historical review of

surveillance test results supports these

conclusions.

3. Involve a significant reduction in a

margin of safety.

The proposed setpoint change evaluated in

Section III results in minimal impact on

system reliability in the interval between

surveillance tests. This is based on the

redundant design of the evaluated systems. A

review of past surveillance history has shown

no evidence of failures which would

significantly impact the reliability of these

systems. Operation of the plant remains

unchanged by this proposed setpoint change.

The assumptions in the Plant Licensing Basis

are not adversely impacted. Therefore, the

proposed changes do not result in a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: February

6, 1998.

Description of amendment request:

The proposed changes would allow

reactor coolant system pressure tests to

be conducted in Cold Shutdown Mode.

Primary containment integrity is not

required in this mode, facilitating

containment access for inspections. The

proposed changes also allow some

outage activities on other systems to

continue during the pressure testing.

The licensee claims the proposed

changes are consistent with the Boiling

Water Reactor Standard Technical

Specifications given in NUREG–1433,

Revision 1.

Basis for proposed no significant

hazards consideration determination:
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19977

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 probability of a leak in the reactor

coolant pressure boundary during reactor

coolant system pressure testing is not

increased by considering the reactor to be in

Cold Shutdown. Since the pressure tests are

performed nearly water solid, at low decay

heat values, and near Cold Shutdown

conditions, the stored energy in the reactor

core will be low. Under these conditions, the

potential for failed fuel and a subsequent

increase in coolant activity is minimized. In

addition, Special Operations LCO [Limiting

Condition for Operation] 3.12.A requires

supporting LCOs for ECCS [emergency core

cooling system]-Cold Condition, Standby Gas

Treatment, Secondary Containment isolation

and Standby Gas Treatment initiation

instrumentation, and Auxiliary Electrical

Systems to be met to ensure secondary

containment integrity is maintained and

capable of handling any airborne

radioactivity or steam leaks that could occur

during the performance of hydrostatic or leak

testing. A listing of secondary containment

isolation valves required to maintain

Secondary Containment Integrity is included

in plant controlled procedures. The required

pressure testing conditions provide adequate

assurance that the consequences of a steam

leak will be conservatively bounded by the

consequences of the postulated main steam

line break outside of primary containment. In

the event of a large primary system leak, the

reactor vessel would rapidly depressurize,

allowing the low pressure core cooling

systems to operate. The capability of these

systems would be adequate to keep the core

flooded under this low decay heat load

condition. Small system leaks would be

detected by leakage inspections before

significant inventory loss occurred.

Therefore, the consequences of an accident

previously evaluated are not significantly

increased.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated.

The proposed changes do not introduce

any new accident initiators or failure

mechanisms since the changes do not involve

any changes to structures, systems, or

components, do not involve any change to

the operation of systems, and alter

procedures only to the extent that the 212°

F limit may be exceeded during reactor

coolant system pressure testing with certain

systems inoperable. There are no alterations

to plant systems designed to mitigate the

consequences of accidents. The only

difference is that a different subset of plant

systems would be utilized for accident

mitigation than those utilized during the Hot

Shutdown Mode. Therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from those

previously evaluated.

3. Involve a significant reduction in the

margin of safety.

Since pressure tests are performed nearly

water solid, at low decay heat values, and

near Cold Shutdown conditions, the stored

energy in the reactor core will be low. Under

these conditions, the potential for failed fuel

and a subsequent increase in coolant activity

is minimized. Since secondary containment

integrity will be maintained, in accordance

with the Special Operations LCO, the

secondary containment will be capable of

handling any airborne radioactivity or steam

leaks that could occur during the

performance of hydrostatic or leak testing.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: February

26, 1998.

Description of amendment request:

The proposed changes would change

the allowed containment leakage rate to

1.5 percent per day, changes the

assumed standby gas treatment system

(SBGT) filter efficiency, and revises

reactor coolant sampling requirements

for low Iodine-131 concentrations.

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 Authority [the licensee] has

evaluated the proposed TS [technical

specification] Amendment and

determined that it does not represent a

significant hazards consideration. Based

on the criteria for defining a significant

hazards consideration established in 10

CFR 50.92, operation of the James A.

FitzPatrick Nuclear Power Plant in

accordance with the proposed

amendment will not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

The proposed changes do not involve a

change to the design or operation of the

plant. The systems affected by this proposed

TS change are not assumed in any safety

analyses to initiate any accident sequence.

Therefore, the probability of any accident

previously evaluated is not increased by this

proposed TS change. The change in the

allowable containment leakage rate (La) is

consistent with the accident analyses. The

assumption of only 90% SBGT filter

efficiency is conservative with actual system

performance and is consistent with

Regulatory Guide 1.52. There is no

significant change to the consequences of an

accident previously evaluated because

maintaining containment leakage within

limits assumed in the accident analyses

ensures that the dose consequences resulting

from an accident are not increased. The

calculated doses with the decreased SBGT

system charcoal efficiency for design basis

accidents are marginally increased but still

meet, and are well below, the dose

acceptance criteria of 10 CFR 100, the SRP

[Standard Review Plan, NUREG–0800], and

GDC [General Design Criterion] 19 of

Appendix A to 10 CFR 50. The proposed TS

changes maintain an equivalent level of

reliability and availability for all affected

systems. The ability of the affected systems

associated with maintaining leak rate

integrity to perform their intended function

is unaffected by the proposed TS changes.

Implementation of these changes will

provide continued assurance that specified

parameters associated with containment

integrity will remain within acceptance

limits, and as such, will not significantly

increase the consequences of a previously

evaluated accident. The change in the value

of .007 [microcurie]/ml to .002 [microcurie]/

ml in section of 4.6.C. ‘‘Coolant Chemistry’’

is a minor editorial change, is more

conservative, and will correct the

inconsistency between the technical

specification and its basis and as such, will

not significantly increase the consequences

of a previously evaluated accident.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed amendment changes the

allowed containment leakage rate to 1.5%,

changes the assumed value for SBGT system

charcoal filter efficiency, and changes a

specification in section of 4.6.C. ‘‘Coolant

Chemistry’’ from the value of .007

[microcurie]/ml to .002 [microcurie]/ml. No

new accident modes are created by clarifying

the numerical value of the allowable

containment leakage rate (La) or changing the

assumed value for the SBGT system charcoal

filter efficiency. No safety-related equipment

or safety functions are altered, or adversely

affected, as a result of these changes. The

proposed changes will not introduce failure

mechanisms beyond those already

considered in the current plant safety

analyses. Changing the allowable leakage

rate, the assumed value for the efficiency of

the SBGT system charcoal filter, and the

specification in the bases section of 4.6.C.

‘‘Coolant Chemistry’’ does not contribute to

the possibility of a new or different kind of

accident or malfunction from those

previously analyzed.
19978 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

(3) Involve a significant reduction in the

margin of safety because:

The proposed amendment changes the

allowed containment leakage rate to 1.5%,

changes the assumed value for SBGT system

charcoal filter efficiency, and changes a

specification in section of 4.6.C. ‘‘Coolant

Chemistry’’ from the value of .007

[microcurie]/ml to .002 [microcurie]/ml. The

design of the FitzPatrick plant is not

changed. The methodology for test

performance is unchanged and Type A, B

and C tests will continue to be performed at

[greater than or equal to] Pa. The value of La

specified in proposed specification 6.20 is

consistent with the accident analyses,

therefore, the dose consequences of any

analyzed accidents are not increased as a

result of this change. The calculated doses as

a result of the decrease in the assumed

efficiency of the SBGT system charcoal filters

for design basis accidents are marginally

increased but still meet, and are well below,

the dose acceptance criteria of 10 CFR 100,

the SRP, and GDC 19 of Appendix A to 10

CFR 50. The change in the specification in

section 4.6.C. ‘‘Coolant Chemistry’’ from .007

[microcurie]/ml to .002 [microcurie]/ml is a

minor editorial change, is more conservative,

and will correct the inconsistency between

the technical specification and its basis.

Therefore, the proposed changes provide

continued assurance of the leak tightness of

the containment and conservatively assume

SBGT system charcoal filter efficiency for the

purpose of dose calculations for design basis

accidents without adversely affecting the

public health and safety and, as such, 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: March

30, 1998.

Description of amendment request:

The proposed changes would change

the interval of selected Logic System

Functional Tests (LSFT) from

semiannually to once per 24 months.

The definition of LSFT is also revised to

be consistent with the Boiling Water

Reactor Standard Technical

Specifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The only significant change proposed by

this application involves the extension of the

surveillance test interval for the LSFTs

required by the TS. The other changes

involve editorial, format, and clarification

changes, which by their nature are of no

safety significance.

Extending the LSFT interval from

semiannually to once per 24 months does not

involve plant physical changes, change any

TS setpoints, or introduce any new mode of

plant operation. Therefore, the change does

not degrade the performance of any safety

system assumed to function in the accident

analysis, and therefore, will not increase the

consequences of an accident.

Extending the LSFT interval from

semiannually to 24 months results in no

significant change in the logic system

unavailability due to equipment failure. The

reliability of safety systems subject to the

LSFT are dominated by that of the

mechanical components, and the logic

system circuit relay coils which are subject

to the more frequent functional test

requirements. These factors are confirmed by

the availability record of the affected safety

system based on the past surveillance test

history. Furthermore, the longer test intervals

reduce the unavailability due to testing for

the applicable safety system while the plant

is operating. For these reasons, there is not

a significant increase in the probability of an

accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not introduce

any new accident initiators or failure

mechanisms since the changes do not

introduce any new modes of plant operation,

make any physical changes, or change any TS

setpoints. The changes reduce the probability

of accidents initiated by test-induced plant

transients by reducing the number of times

the tests must be performed.

3. Involve a significant reduction in a

margin of safety.

The proposed changes do not alter the

manner in which safety limits, limiting safety

system settings, or limiting conditions for

operation are determined. In several aspects,

the proposed changes may actually enhance

the margin of safety by reducing the potential

for test-induced plant transients, reducing

the unavailability due to test of the

applicable safety system, and reducing any

potential incremental logic system

component wear. For these reasons, the

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of amendment request: March

26, 1998.

Description of amendment request:

The proposed amendment would revise

Technical Specification 3.1.3.3, ‘‘Rod

Drop Time,’’ to change the applicability

from Mode 3 (hot shutdown) to Modes

1 and 2 (startup and power operation).

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 the Technical

Specification Mode applicability provides

consistency between the testing requirements

as stated in the surveillance requirement of

the Technical Specifications and intended by

the initial conditions specified in the limiting

condition for operations. The proposed

change does not introduce any physical

changes to the plant or equipment already in

place in the plant, the proposed change

ensures that testing of the rod drop times is

performed in a manner that is consistent with

the Technical Specifications and the

assumptions made in the Salem accident

analysis.

Therefore, the proposed amendment does

not increase the probability or consequences

of any 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 a

new component or changes the manner in

which the facility is operated, maintained or

tested. Thus no new accident scenarios,

failure mechanisms or limiting single failures

are introduced as a result of the proposed

change to the facility.

Therefore the proposed amendment does

not create the possibility of a new or different

kind of accident from any previously

evaluated.
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19979

3. The proposed changes does not involve

a significant reduction in a margin of safety.

As stated in question number 2, the

proposed change does not introduce a new

component or changes the manner in which

the facility is operated. Operation of the

facility in accordance with the proposed

amendment would not involve a significant

reduction in the margin of safety. The

Technical Specifications remain the same, as

the input, or initial conditions, of the safety

analysis have not changed. Therefore, there

is no reduction in the margin to 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.

Tennessee Valley Authority, Docket

Nos. 50–260 and 50–296, Browns Ferry

Nuclear Plant, Units 2, and 3, Limestone

County, Alabama

Date of amendment request: March 3,

1998.

Description of amendment request:

The proposed amendment would

change the Browns Ferry Nuclear Plant

Unit 2 and Unit 3 Technical

Specification Figure 3.6–1 which

contains the reactor vessel pressuretemperature (PT) limits. The change

would extend the validity of the curves

to 32 effective full-power years (EFPY).

The current PT curves are effective up

to 12 EFPY. In addition to revised PT

curves, several changes to the notes

applicable to the curves are also

proposed to be consistent with the

supporting analysis.

The proposed PT curves also would

support a planned 5% power increase

for each unit. Approval of the proposed

power increase is pending and is the

subject of a separate action before the

Commission.

The Tennessee Valley Authority has

submitted the proposed change in

current technical specification (CTS)

format and in the improved standard

technical specification (ISTS) format.

Conversion to the ISTS format is

pending.

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.

A. The changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed changes

merely adjust the reference temperature for

the limiting reactor vessel beltline material to

account for accumulated and projected

irradiation effects. The adjusted reference

temperature analyses were performed in

accordance with the requirements of

Appendix G of 10 CFR part 50 and the

guidance contained in Regulatory Guide 1.99,

Revision 2. The changes do not otherwise

affect the manner by which the facility is

operated and do not change any facility

design feature or equipment. Since the

protection previously provided will continue

to be provided and there is no change to the

facility or operating procedures, there is no

effect upon the probability or consequences

of any accident previously analyzed.

B. The changes do not create the possibility

of a new or different kind of accident from

any accident previously evaluated (10 CFR

50.92(c)(2)) because no new failure modes are

introduced. The proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated (10 CFR 50.92(c)(2))

because they do not affect the function of any

facility structure, system or component, or

affect the manner by which the facility is

operated.

C. The changes do not involve a significant

reduction in a margin of safety (10 CFR

50.92(c)(3)) because the proposed changes

assure that the reactor vessel PT limits will

be valid for operation up to 32 EFPY and that

the safety margins specified in Appendix G

of 10 CFR part 50 will be maintained.

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: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

ET 10H 400 West Summit Hill Drive,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

February 13, 1998 (TS 97–03).

Brief description of amendments: The

proposed amendments would change

the Sequoyah (SQN) Technical

Specification (TS) by adding a new

limiting condition for operation (LCO)

that addresses requirements for the

main feedwater isolation, regulating,

and bypass valves.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has concluded that operation of SQN

Units 1 and 2 in accordance with the

proposed change to the TSs [or operating

license(s)], does not involve a significant

hazards consideration. TVA’s conclusion is

based on its evaluation, in accordance with

10 CFR 50.91(a)(1), of the three standards set

forth in 10 CFR 50.92(c).

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

TVA will not change plant components,

functions, or operating practices by

implementing a change that adds a TS

requirement for the main feedwater isolation,

regulating, and bypass valves. TVA will

maintain and verify operability of these

valves through the proposed surveillance and

actions to ensure the accident mitigation

functions are available when applicable.

These valves are not considered to be the

source of an accident and the conservative

addition of a requirement to maintain their

safety function will not increase the

probability of an accident. TVA will not

increase the consequences of an accident by

implementing this change because this

addition ensures that the isolation of main

feedwater is available to mitigate the

consequences of an accident.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

TVA will not alter plant equipment or

operating activities in the implementation of

the proposed TS change. The valves used for

the isolation of main feedwater are not a

potential source for accidents and are

designed for accident mitigation purposes.

Therefore, TVA will not create the possibility

of an accident of a different kind.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

TVA maintains and ensures the availability

of the isolation function for the main

feedwater system as assumed in the SQN

accident analysis. TVA proposes this TS

change to further assure this capability and

to meet the requirements of 10 CFR 50.36.

TVA will not change the methods of

operating the plant or setpoints associated

with safety-related equipment in the

implementation of this request. Therefore,

TVA will not reduce the margin of safety by

implementing a TS LCO for the isolation

functions of the main feedwater system.

The NRC 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.
19980 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

February 13, 1998 (TS 97–07).

Brief description of amendments: The

proposed amendments would change

the Sequoyah (SQN) Technical

Specification (TS) requirements for

main steam isolation valves (MSIVs) to

incorporate MSIV requirements

consistent with the Westinghouse

Standard TS (NUREG–1431) and would

add testing requirements for the MSIVs

that ensure the valves close on an actual

or simulated automatic actuation signal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has concluded that operation of SQN

Units 1 and 2, in accordance with the

proposed change to the TSs, does not involve

a significant hazards consideration. TVA’s

conclusion is based on its evaluation, in

accordance with 10 CFR 50.91(a)(1), of the

three standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes provide

enhancements and clarifications of the

requirements for inoperable MSIVs and

periodic testing provisions. These changes do

not alter the safety functions of the MSIVs or

the operating practices that govern their

application to plant conditions. The actions

for Modes 2 and 3 are revised such that a

longer time could occur before an inoperable

MSIV is closed or the unit is placed in a

mode that does not apply. However, this

increase will not significantly impact the

ability of the valves to mitigate an accident

or affect the accident generation possibility.

This is based on the low probability of an

accident occurring that would require closure

of the MSIVs and reasonable time intervals

to transition to lower modes based on

operating experience to reach the required

modes in an orderly manner without

challenging unit systems.

The MSIVs provide accident mitigation

functions but do not contribute to accident

generation. The MSIV functions have not

been altered by the proposed changes.

Therefore, the proposed changes will not

increase the probability of a previously

evaluated accident. Based on the above

discussions, the proposed changes will not

significantly increase the consequences of an

accident and in some instances they will

enhance the safety functions.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The primary function of the MSIVs is to

support accident mitigation and are not a

significant contributor to events that could

generate accidents. The main impact that

could result from an inoperable MSIV is an

inadvertent closure that results in a unit trip.

This event is bounded by the accidents that

are currently evaluated for SQN. Since the

proposed change does not alter MSIV

functions and the new surveillance will be

performed in modes that will not challenge

unit systems, the possibility of a new or

different kind of accident is not created.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed changes clarify and enhance

the current SQN requirements for the MSIVs

with one exception. This exception is the

completion time added to the Modes 2 and

3 action that could be a negative impact to

the margin of safety. This change could allow

the MSIV safety function to be inoperable for

a longer period of time. The overall effect of

the proposed changes considering the

additional end-device testing, periodic

verification of inoperable MSIV closure, and

removal of the action to allow MSIV closure

in Mode 1, is considered a positive impact

to the margin of safety. Therefore, there is not

a significant reduction in the margin of

safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Virginia Electric and Power Company,

Docket Nos. 50–338 and 50–339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request: March

25, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) Sections 6.1.1;

6.2.1.b; 6.5.1.1; 6.5.1.6. a, d, h, and m;

6.5.1.7.c; 6.5.1.8; 6.14.1.2; 6.15.b;

6.2.3.5; 6.5.1.2; 6.5.1.7.a for Unit 1 and

6.1.1; 6.2.1.b; 6.5.1.1; 6.5.1.6. a, d, h,

and m; 6.5.1.7.c; 6.5.1.8; 6.13.b; 6.14.b;

6.2.3.5; 6.5.1.2; and 6.5.1.7.a for Unit 2,

changing the title of Station Manager to

Site Vice President, and the titles of the

Assistant Station Managers to ManagerStation Operation 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: The Alderman Library, Special
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19981

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: P. T. Kuo,

Acting Project Director.

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:

November 5, 1997.

Description of amendment request:

The proposed Operating License change

and changes to the technical

specifications (TS) would permit the use

of a temporary alternate supply line

(jumper) to provide service water (SW)

to the component cooling heat

exchangers. The temporary jumper will

permit maintenance to be performed on

the existing supply line.

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 changes against the

criteria of 10 CFR 50.92 and has concluded

that the changes do not pose a significant

safety hazards consideration as defined

therein. The proposed Operating License and

Technical Specifications and Bases changes

are necessary to allow the use of a temporary,

seismic, non-missile protected jumper to

provide service water (SW) to the Component

Cooling Heat Exchangers (CCHXs) while

maintenance work is performed on the

existing SW supply line to the CCHXs. Since

there is only one SW supply line to the

CCHXs, an alternate SW supply must be

provided whenever the line is removed from

service. The temporary jumper provides this

function.

The use of the temporary jumper has been

thoroughly evaluated, and appropriate

constraints and compensatory measures

(including a Contingency Action Plan) have

been developed to ensure that the temporary

jumper is reliable, safe, and suitable for its

intended purpose. A complete and

immediate loss of SW supply to the operating

CCHXs is not considered credible, given the

project constraints and the unlikely

probability of a generated missile. Existing

station abnormal procedures already address

a loss of component cooling, and the use of

alternate cooling for a loss of decay heat

removal, in the unlikely event that they are

required. Furthermore, appropriate mitigative

measures have been identified to address

potential flooding concerns. The minor

administrative changes merely correct a table

format inconsistency and update Basis

section references.

Consequently, the operation of Surry

Power Station with the proposed amendment

and license condition will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The SW and CC Systems will function as

designed under the Unit operating

constraints specified by this project (i.e., Unit

2 in operation and Unit 1 in a refueling

outage), and the potential for a loss of

component cooling is already addressed by

Station Abnormal Procedures. Therefore,

there is no increase in the probability of an

accident previously evaluated. The

possibility of flooding due to failure of the

temporary SW supply jumper in the Turbine

Building basement has been evaluated and

dispositioned by the implementation of

appropriate precautions and compensatory

measures to preclude damage to the

temporary jumper and to respond to a

postulated flooding event. A flood watch will

be present around-the-clock with authority

and procedural guidance to isolate the

jumper, if required. Furthermore, the CCHXs

serve no design basis accident mitigating

function. Therefore, the consequences of an

accident previously evaluated are not

increased.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The SW and CC Systems’ design

functions and basic configurations are

not being altered as a result of using a

temporary SW supply jumper. The

temporary jumper is designed to be

safety-related and seismic with all of the

design attributes of the normal SW

supply line, except for the automatic

isolation function and complete missile

protection. The design functions of the

SW and CC systems are unchanged as a

result of the proposed changes due to (1)

required plant conditions, (2)

compensatory measures, (3) a

Contingency Action Plan for restoration

of the normal SW supply if required,

and (4) strict administrative control of

the temporary SW valve to preclude

flooding or to isolate non-essential SW

within the design basis assumed time

limits. Unit 1 will be in a plant

condition which will provide adequate

time to restore the normal SW supply,

if required. Therefore, since the SW and

CC systems will basically function as

designed and will be operated in their

basic configuration, the possibility of a

new or different type of accident than

previously evaluated in the UFSAR

[Updated Final Safety Analysis Report]

is not created.

3. Involve a significant reduction in a

margin of safety.

The margin of safety as defined in the

Technical Specifications is not reduced

since an operable SW flowpath to the

required number of CCHXs is provided,

and Unit operating constraints,

compensatory measures and

contingencies will be implemented as

required to ensure the integrity and the

capability of the SW flowpath. The use

of the temporary jumper will be limited

to the time period when missile

producing weather is not expected, and

Unit 1 meets specified unit conditions.

Therefore, the temporary SW jumper,

under the imposed project constraints

and compensatory measures, provides

the same reliability as the normal SW

supply line. Furthermore, the

Probabilistic Safety Assessment for

Surry Power Station has been reviewed

relative to potential flooding when the

temporary SW jumper is in use. It has

been determined that due to the SW

restoration project’s compensatory and

contingency measures, as well as the

constraints imposed by the Maintenance

Rule online risk matrix, the impact on

core damage frequency due to flooding

is negligible.

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.

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.
19982 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Northern States Power Company,

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

Island Nuclear Generating Plant, Units

1 and 2, Goodhue County, Minnesota

Date of amendment request: March 6,

1998.

Description of amendment request:

The proposed amendment would (1)

update the Technical Specification

heatup and cooldown rate curves and

extend their reactor fluence limit from

the current 20 effective full power years

(EFPY) to a new value of 35 EFPY, (2)

incorporate into Technical

Specifications the use of a Pressure and

Temperature Limits Report (PTLR), and

(3) change the power-operated relief

valves (PORVs) temperature

requirement for operability.

Date of individual notice in the

Federal Register: March 27, 1998 (63 FR

14972).

Expiration date of individual notice:

April 27, 1998.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Commonwealth Edison Company,

Docket No. 50–237, Dresden Nuclear

Power Station, Unit 2, Grundy County,

Illinois

Date of application for amendment:

March 19, 1998, as supplemented by

letters dated March 28, 1998, and April

3, 1998.

Brief description of amendment: The

amendment changes the Technical

Specifications (TS) by revising the

Dresden, Unit 2, Minimum Critical

Power Ratio (MCPR) in TS Section 2.1.B

and footnotes in TS Section 5.3.A, to

allow the use of Siemens Power

Corporation ATRIUM–9B fuel for all

operating Modes at Dresden, Unit 2,

Cycle 16.

Date of issuance: April 10, 1998.

Effective date: Immediately, to be

implemented within 30 days.

Amendment No.: 168.

Facility Operating License No. DPR–

19: The amendment revised the TS.

Public comments requested as to

proposed no significant hazards

consideration: Yes (63 FR 14735 dated

March 26, 1998). The notice provided

an opportunity to submit comments on

the Commission’s proposed no

significant hazards consideration

determination. No comments have been

received. The notice also provided for

an opportunity to request a hearing by

April 27, 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 March 28, 1998, and April 3,

1998, letters provided additional

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.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant Middlesex County,

Connecticut

Date of application for amendment:

Two applications, both dated May 30,

1997.

Brief description of amendment:

Changes Administrative Controls

Section of the Technical Specifications

to implement new Certified Fuel

Handler position and to implement

revised management responsibilities

and titles that reflect the permanently

shut down status of the plant. In

addition, minor typographical errors

were corrected.

Date of issuance: March 27, 1998.

Effective date: Date of issuance, but to

be implemented within 60 days of

issuance.

Amendment No.: 192.

Operating License No. DPR–61:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38132 and

62 FR 38133). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 27, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

March 26, 1997.

Brief description of amendment: The

amendment revises the Containment

Systems technical specifications (TS) to

incorporate a note to allow opening an

operable airlock door to perform repairs

on inoperable airlock components when

the other airlock door is inoperable.

This amendment is in partial response

to Consumers Energy’s March 26, 1997,

application. The Consumers Energy

request also proposed revising the

requirements contained in TS sections

3.6 and 4.5 to closely emulate the format

and content of NUREG–1432, ‘‘Standard

Technical Specifications, Combustion

Engineering Plants,’’ (STS). That portion

of the Consumers Energy request

remains under staff review and will be

addressed in a separate evaluation.

Date of issuance: April 8, 1998.

Effective date: April 8, 1998.

Amendment No.: 179.

Facility Operating License No. DPR–

20. Amendment revised the Technical

Specifications.
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19983

Date of initial notice in Federal

Register: December 17, 1997 (62 FR

66136).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

January 28, 1998 (NRC–98–0008), as

supplemented on March 10, 1998 (NRC–

98–0036).

Brief description of amendment: The

amendment revises the technical

specifications (TSs) by modifying the

‘‘#’’ footnote to Table 1.2 and the ‘‘*’’

footnote to surveillance requirements

4.9.1.2 and 4.9.1.3 to permit the Reactor

Mode Switch to be placed in the Run or

Startup/Hot Standby positions to test

switch interlock functions provided that

all control rods are verified to remain

fully inserted in core cells containing

one or more fuel assemblies.

Date of issuance: March 31, 1998.

Effective date: March 31, 1998, with

full implementation within 90 days.

Amendment No.: 116.

Facility Operating License No. NPF–

43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 25, 1998 (63 FR

9599) The March 10, 1998, supplement

requested a change in the

implementation period and was not

outside the scope of the initial proposed

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

March 31, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Duke Energy Corporation (DEC), Docket

Nos. 50–369 and 50–370, McGuire

Nuclear Station, Units 1 and 2,

Mecklenburg County, North Carolina

Date of application for amendments:

December 17, 1997.

Brief description of amendments: The

amendments revise Technical

Specification Section 6.9.1.9 to

reference updated or recently approved

topical reports, which contain

methodologies used to calculate cyclespecific limits contained in the Core

Operating Limits Report. For several

reports DEC indicated staff approval,

but neglected to provide an ‘‘A’’

designation for the report number. Upon

agreement by DEC, the staff has made

these appropriate editorial corrections.

These topical reports have all been

previously approved by the staff under

licensing actions separate from the

current amendment request.

Date of issuance: April 8, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1—178; Unit

2—160.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4311).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1,

Pope County, Arkansas

Date of amendment request: February

9, 1998.

Brief description of amendment: The

amendment approves the use of the

repair roll technology (reroll) for the

upper tubesheet region of the ANO–1

steam generators. The reroll technology

is an alternative to the either sleeving or

plugging steam generator tubes found

during inservice inspections to have

defects that exceed the stated repair

criteria. The reroll methodology works

by creating a new mechanical tube to

tubesheet structural joint below the tube

defect indication.

Date of issuance: April 10, 1998.

Effective date: April 10, 1998.

Amendment No.: 190.

Facility Operating License No. DPR–

51: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 24, 1998 (63 FR

9268).

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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1,

Pope County, Arkansas

Date of amendment request: April 1,

1998, as supplemented by letter dated

April 8, 1998.

Brief description of amendment: The

amendment allows approximately 440

steam generator tubes with confirmed

volumetric indications within the upper

tube sheet to remain in service during

Cycle 15. The amendment revises TS

4.18.5.b to incorporate five criteria

which need to be satisfied to allow

steam generator tubes to remain in

service during Cycle 15 with indications

of outer diameter intergranular attack

(ODIGA) in the upper tube sheet region

of the steam generators.

Date of issuance: April 10, 1998.

Effective date: April 10, 1998.

Amendment No.: 191.

Facility Operating License No. DPR–

51: The amendment revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated April 10, 1998.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request:

November 18, 1996, as supplemented by

letter dated January 21, 1998.

Brief description of amendment: The

amendment changes Technical

Specification (TS) Surveillance

Requirement 4.4.8.3.1.b to test the

Shutdown Cooling System suction line

relief valves in accordance with TS

4.0.5. Editorial changes to 4.4.8.3.1 and

4.4.8.3.1.a have also been made.

Date of issuance: April 1, 1998.

Effective date: April 1, 1998.

Amendment No.: 140.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 11, 1998 (63 FR

6985).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 1, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans
19984 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 24,

1997, as supplemented by letter dated

January 21, 1998.

Brief description of amendment: The

amendment modifies Technical

Specifications (TS) 3.1.1.1, 3.1.1.2,

3.10.1 and Figure 3.1–1 by removing

cycle dependent boron concentration

and boration flow rate from the Action

Statements and removing the ‘‘RWSP at

1720 ppm’’ curve from the figure. A

change to TS Bases 3/4.1.1.1 and 3/

4.1.1.2 has been included to support

this change.

Date of issuance: April 8, 1998.

Effective date: April 8, 1998, to be

implemented within 60 days.

Amendment No.: 141.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33123).

The January 21, 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 8, 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.

Entergy Operations, Inc., System Energy

Resources, Inc., South Mississippi

Electric Power Association, and Entergy

Mississippi, Inc.

Docket No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of application for amendment:

October 28, 1997, as supplemented by

letter dated January 9, 1998.

Brief description of amendment: The

amendment revised the Technical

Specifications (TSs) to implement the

containment leak rate testing provisions

of 10 CFR Part 50, Appendix J, Option

B.

Date of issuance: April 6, 1998.

Effective date: April 6, 1998.

Amendment No: 135.

Facility Operating License No. NPF–

29: Amendment revises the TSs.

Date of initial notice in Federal

Register: December 3, 1997 ( 62 FR

63976).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 6, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

IES Utilities Inc, Central Iowa Power

Cooperative, and Corn Belt Power

Cooperative, Docket No. 50–331, Duane

Arnold Energy Center, Linn County,

Iowa

Date of application for amendment:

February 3, 1998.

Brief description of amendment: The

amendment changes the operability

requirement for the Standby Liquid

Control system to Run/Power

Operations and Startup.

Date of issuance: March 31, 1998.

Effective date: March 31, 1998.

Amendment No.: 221.

Facility Operating License No. DPR–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 26, 1998 (63 FR

9874).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, Iowa

52401.

IES Utilities Inc, Central Iowa Power

Cooperative, and Corn Belt Power

Cooperative, Docket No. 50–331, Duane

Arnold Energy Center, Linn County,

Iowa

Date of application for amendment:

October 3, 1997 as supplemented on

December 10, 1997.

Brief description of amendment: The

amendment revises the Operating

License to allow the start of core offload

as soon as 60 hours after shutdown.

Date of issuance: April 2, 1998.

Effective date: April 2, 1998.

Amendment No.: 222.

Facility Operating License No. DPR–

49: Amendment revised the Operating

License.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4314).

The December 10, 1997 submittal

provided clarifying information that did

not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, Iowa

52401.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

October 20, 1997, as supplemented

February 10, and March 10, 1998.

Brief description of amendment: The

amendment replaces in their entirety

the existing Technical Specifications

incorporated in Facility Operating

License No. DPR–36 as Appendix A.

Maine Yankee developed the revised

Technical Specifications, titled

Permanently Defueled Technical

Specifications, to reflect the

permanently shutdown and defueled

status of the plant. Changes were made

to the definitions, limiting conditions

for operation, surveillance, and

administrative control sections.

Date of issuance: March 30, 1998.

Effective date: March 30, 1998.

Amendment No.: 161.

Facility Operating License No. DPR–

36: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 3, 1997 (62 FR

63978). The February 10, and March 10,

1998, submittals added additional

programs to the Section 5.5 Procedures

and Section 5.6 Programs and Manuals

did not change the proposed no

significant hazards determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated March 30, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of application for amendment:

October 31, 1997, as supplemented by

letter dated February 3, 1998.

Brief description of amendment: This

amendment changes Technical

Specifications to support design

changes to upgrade the analog-based

average power range monitor system

with General Electric’s Nuclear

Measurement Analysis and Control

Power Range Neutron Monitor System,

including an Oscillation Power Range

Monitor function.

Date of issuance: March 31, 1998.

Effective date: As of the date of

issuance to be implemented upon
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19985

completion and acceptance of design

modifications resulting from the

installation of the Nuclear Measurement

Analysis and Control Power Range

Neutron Monitor System.

Amendment No.: 80.

Facility Operating License No. NPF–

69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68310).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 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.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

September 26, 1997, as supplemented

by letter dated March 12, 1998.

Description of amendment request:

The amendment revises Technical

Specification 3.7.6, ‘‘Control Room

Emergency Makeup Air and Filtration,’’

and its associated Bases to separate the

requirements for the control room air

conditioning subsystem from the

requirements for control room makeup

air and filtration subsystem based on

system function. The amendment also

increases the allowed outage time for

the Control Room Air Conditioning

Subsystem.

Date of issuance: April 9, 1998.

Effective date: As of the date of

issuance, with full implementation

within 60 days.

Amendment No.: 56.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54874). The March 12, 1998,

supplemental letter 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 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–336,

Millstone Nuclear Power Station, Unit

No. 2, New London County, Connecticut

Date of application for amendment:

December 1, 1997.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by adding a 2.0

second plus or minus 0.1 second time

delay to the 4.16 kV Emergency Bus

Undervoltage Loss of Power, Level One,

trip setpoint and allowable values in TS

Table 3.3–4.

Date of issuance: April 1, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 214.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 14,1998 (63 FR 2280).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 1, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

locations: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of application for amendment:

November 11, 1997.

Brief description of amendment: The

amendment allows NNECO to credit

soluble boron for maintaining keffective at less than or equal to 0.95

within the spent fuel pool rack matrix

following a seismic event of a

magnitude greater than or equal to an

operating basis earthquake.

Date of issuance: April 9, 1998.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 158.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 3, 1997 (62 FR

63980).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

locations: 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.

Pacific Gas and Electric Company,

Docket No. 50–133, Humboldt Bay

Power Plant, Unit 3, Humboldt County,

California

Date of application for amendment:

December 9, 1996, as supplemented on

June 12, 1997, and March 13, 1998.

Brief description of amendment: This

amendment revised the Technical

Specification to incorporate the

requirements of appendix I of 10 CFR

Part 50, into the Radiological Effluent

Technical Specification (RETS) and to

relocate the controls and limitations on

RETS and radiological environmental

monitoring (Currently in the Technical

Specifications) to the Offsite Dose

Calculation Manual and the Process

Control Program. The amendment also

revised the Technical Specifications to

implement Generic Letter 89–10 (GL

89–10) and to incorporate the

requirements of the revised 10 CFR Part

20.

Date of issuance: April 8, 1998.

Effective date: As of the date of

issuance and shall be implemented no

later than 30 days from the date of

issuance.

Amendment No.: 32.

Facility Operating License No. DPR–7:

Amendment revised the TS.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18174).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Humboldt County Library, 131

3rd Street, Eureka, California 95501.

Pennsylvania Power and Light

Company, Docket No. 50–387,

Susquehanna Steam Electric Station,

Unit 1, Luzerne County, Pennsylvania

Date of application for amendment:

August 26, 1997, as supplemented by

letters dated December 4, 1997,

February 2, March 19, and April 2,

1998.

Brief description of amendment: This

amendment changes the Susquehanna

Unit 1 Technical Specifications to

support the use of the Siemens Power

Corporation ATRIUM–10 fuel design in

the upcoming Cycle 11 refueling outage.

Date of issuance: April 6, 1998.

Effective date: As of date of issuance,

to be implemented within 30 days.
19986 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

Amendment No.: 174.

Facility Operating License No. NPF–

14: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68314).

The December 4, 1997, February 2,

March 19, and April 2, 1998, submittals

provided clarifying information that did

not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 6, 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:

September 2, 1997.

Brief description of amendment:

These amendments revise LGS, Units 1

and 2, TS Section 4.0.5, and Bases

Sections B 4.0.5 and B 3/4.4.8 regarding

the surveillance requirements associated

with Inservice Testing and Inservice

Inspection Programs of the American

Society of Mechanical Engineers Code

Class 1, 2, and 3 components.

Date of issuance: March 31, 1998.

Effective date: March 31, 1998.

Amendment Nos.: 125 and 89.

Facility Operating License No. NPF–

39: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 11, 1998 (63 FR

6990).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, Alabama

Date of amendments request: July 23,

1997, as supplemented by letters dated

September 30, October 27, and

December 18, 1997, and February 12,

1998.

Brief Description of amendments: The

amendments revise the PressureTemperature Limit Heatup, Cooldown,

and Hydrostatic Testing curves for

Farley Units 1 and 2 and relocate the

curves from the Technical

Specifications to a Pressure and

Temperature Limits Report for each

unit.

Date of issuance: April 9, 1998.

Effective date: As of the date of

issuance to be implemented for Unit 1

prior to entering Mode 4 for Cycle 16

refueling outage (fall 1998); for Unit 2

prior to entering Mode 4 for Cycle 13

refueling outage (spring 1998).

Amendment Nos.: Unit 1–136; Unit

2–128.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: September 10, 1997 (62 FR

47699); January 14, 1998 (63 FR 2281);

February 23, 1998 (63 FR 9020).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama.

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 request: October

24, 1997 (TXX–97228).

Brief description of amendments: The

amendments revise core safety limit

curves and Overtemperature N–16

reactor trip setpoints based on analyses

of the core configuration and expected

operation for CPSES Unit 1, Cycle 7.

The changes apply equally to CPSES

Units 1 and 2 licenses since the

Technical Specifications are combined.

Date of issuance: March 27, 1998.

Effective date: March 27, 1998, to be

implemented within 30 days.

Amendment Nos.: Unit 1—

Amendment No. 57; Unit 2—

Amendment No. 43.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications/

operating licenses.

Date of initial notice in Federal

Register: November 19, 1997 (62 FR

61847).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 27, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

August 8, 1997, as supplemented by

letter dated January 16, 1998.

Brief description of amendment: The

amendment revises the Callaway Plant,

Unit 1 Technical Specification Table

3.3–3 Functional Units 4.b.2 and 5.a.2 to

make the number of main steam and

feedwater isolation system (MSFIS)

channels consistent with the solid state

protection system, adds a clarifying note

and changes Table 4.3–2 Functional

Units 4.b.2 and 5.a.2 slave relay

quarterly test to a monthly staggered

actuation logic test.

Date of issuance: March 25, 1998.

Effective date: March 25, 1998, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 123.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 17, 1997 (62 FR

66143) The January 16, 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 March 25, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of MissouriColumbia, Elmer Ellis Library,

Columbia, Missouri 65201–5149.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

October 17, 1997, as supplemented by

letters dated March 3, 1998, and March

17, 1998.

Brief description of amendment: The

amendment revises the technical

specifications to modify the heatup and

cooldown curves and the maximum

allowable power operated relief valve

setpoint curves for cold overpressure

protection.

Date of issuance: April 2, 1998.

Effective date: April 2, 1998, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 124.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 14, 1998 (63 FR 2282).

The March 3, 1998, and March 17,

1998, supplemental letters provided
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19987

additional clarifying information and

did not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of MissouriColumbia, 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:

October 11, 1996.

Brief description of amendment: The

amendment revises the Technical

Specifications regarding the amount of

foam concentrate required to support

operability of the reactor recirculation

motor generator set foam fire

suppression system.

Date of Issuance: March 31, 1998.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 156.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54877).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated March 31, 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:

June 9, 1997.

Brief description of amendment: The

amendment revises Technical

Specification Section 6.0 to add and

revise reference to NRC-approved

methodologies which will be used to

validate or generate the cycle-specific

thermal hydraulic stability based

operating limits in the Vermont Yankee

Core Operating Limits Report.

Date of Issuance: April 7, 1998.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 157.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR

43377).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated April 7, 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:

August 22, 1996.

Brief description of amendment: The

amendment revises the Technical

Specifications to change the action

statement for the high range stack noble

gas monitor based on the guidance of

Generic Letter 83–36, NUREG–0737

Technical Specifications.

Date of Issuance: April 8, 1998.

Effective date: April 8, 1998, to be

implemented within 30 days.

Amendment No.: 158.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30647).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated April 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: January

28, 1998.

Brief description of amendment: The

amendment revises Technical

Specification §§ 6.3 and 6.12 to reflect a

merger for the positions of

Superintendent Radiation Protection

and Superintendent Chemistry into one

new position, Manager Chemistry/

Radiation Protection.

Date of issuance: March 30, 1998.

Effective date: March 30, 1998.

Amendment No.: 115.

Facility Operating License No. NPF–

42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 25, 1998 (63 FR

9614).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 30, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

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 for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an
19988 Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective 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 application for

amendment, (2) the amendment to

Facility Operating License, 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 room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By May

22, 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. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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 the factors specified

in 10 CFR 2.714(a)(1)(i)–(v) and

2.714(d).

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: April 2,

1998 (NRC–98–0062).
Federal Register / Vol. 63, No. 77 /Wednesday, April 22, 1998 / Notices 19989

Description of amendment request:

The amendment revised the action

specified in Technical Specification

Table 3.3.7.5–1 if one channel of

drywell oxygen monitoring is

inoperable.

Date of issuance: April 3, 1998.

Effective date: April 3, 1998, with full

implementation by April 6, 1998.

Amendment No.: 117.

Facility Operating License No. NPF–

43: Amendment revises the License and

the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No. The Commission’s

related evaluation of the amendment,

finding of emergency circumstances,

and final determination of no significant

hazards consideration are contained in

a Safety Evaluation dated April 3, 1998.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Project Director: Cynthia A.

Carpenter.

Dated at Rockville, Maryland, this 15th day

of April 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–10470 Filed 4–21–98; 8:45 am]

BILLING CODE 7590–01–P

RAILROAD RETIREMENT BOARD

Proposed Collection; Comment

Request

SUMMARY: In accordance with the

requirement of Section 3506 (c)(2)(A) of

the Paperwork Reduction Act of 1995

which provides opportunity for public

comment on new or revised data

collections, the Railroad Retirement

Board (RRB) will publish periodic

summaries of proposed data collections.

Comments are invited on: (a) Whether

the proposed information collection is

necessary for the proper performance of

the functions of the agency, including

whether the information has practical

utility; (b) the accuracy of the RRB’s

estimate of the burden of the collection

of the information; (c) ways to enhance

the quality, utility, and clarity of the

information to be collected; and (d)

ways to minimize the burden related to

the collection of information on

respondents, including the use of

automated collection techniques or

other forms of information technology.

Title and purpose of information

collection: Employer Service and

Compensation Reports; OMB 3220–0070

Section 2(c) of the Railroad

Unemployment Insurance Act (RUIA)

specifies the maximum normal

unemployment and sickness benefits

that may be paid in a benefit year.

Section 2(c) further provides for

extended benefits for certain employees

and for beginning a benefit year early for

other employees. The conditions for

these actions are prescribed in 20 CFR

302.

All information about creditable

railroad service and compensation

needed by the RRB to administer

Section 2(c) is not always available from

annual reports filed by railroad

employers with the RRB (OMB 3220–

0008). When this occurs, the RRB must

obtain supplemental information about

service and compensation.

The RRB utilizes Form UI–41,

Supplemental Report of Service and

Compensation, and Form UI–41a,

Supplemental Report of Compensation,

to obtain the additional information

about service and compensation from

railroad employers. Completion of the

forms is mandatory. One response is

required of each respondent. The RRB

proposes to revise Form UI–41 and UI–

41a to add language required by the

Paperwork Reduction Act of 1995.

Minor editorial changes are also

proposed. The completion time for

Form UI–41 and UI–41a is estimated at

8 minutes per response.

ADDITIONAL INFORMATION OR COMMENTS:

To request more information or to

obtain a copy of the information

collection justification, forms, and/or

supporting material, please call the RRB

Clearance Officer at (312) 751–3363.

Comments regarding the information

collection should be addressed to

Ronald J. Hodapp, Railroad Retirement

Board, 844 N. Rush Street, Chicago,

Illinois 60611–2092. Written comments

should be received on or before June 22,

1998.

Chuck Mierzwa,

Clearance Officer.

[FR Doc. 98–10586 Filed 4–21–98; 8:45 am]

BILLING CODE 7905–01–M

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 35–26857]

Filings Under the Public Utility Holding

Company Act of 1935, as amended

(‘‘Act’’)

April 16, 1998.

Notice is hereby given that the

following filing(s) has/have been made

with the Commission pursuant to

provisions of the Act and rules

promulgated under the Act. All

interested persons are referred to the

application(s) and/or declaration(s) for

complete statements of the proposed

transaction(s) summarized below. The

application(s) and/or declaration(s) and

any amendments is/are available for

public inspection through the

Commission’s Office of Public

Reference.

Interested persons wishing to

comment or request a hearing on the

application(s) and/or declaration(s)

should submit their views in writing by

May 12, 1998, to the Secretary,

Securities and Exchange Commission,

Washington, D.C. 20549, and serve a

copy on the relevant applicant(s) and/or

declarant(s) at the address(es) specified

below. Proof of service (by affidavit or,

in case of an attorney at law, by

certificate) should be filed with the

request. Any request for hearing should

identify specifically the issues of fact or

law that are disputed. A person who so

requests will be notified of any hearing,

if ordered, and will receive a copy of

any notice or order issued in the matter.

After May 12, 1998, the application(s)

and/or declaration(s), as filed or as

amended, may be granted and/or

permitted to become effective.

American Electric Power Company, Inc.

and Central and South West Corporation (70–

9169).

Notice of Proposal to Amend Certificate of

Incorporation to Increase Number of

Authorized Shares of Common Stock; Order

Authorizing Solicitation of Proxies.

American Electric Power Company,

Inc. (‘‘AEP’’), 1 Riverside Plaza,

Columbus, Ohio 43215, and Central and

South West Corporation (‘‘CSW’’), 1616

Woodall Rodgers Freeway, Dallas, Texas

75266, each a registered holding

company, have filed a joint declaration

with this Commission under sections

6(a)(2), 7 and 12(e) of the Public Utility

Holding Company Act of 1935, as

amended (‘‘Act’’), and rules 62 and 65

under the Act.

AEP and CSW have entered into an

Agreement and Plan of Merger, dated as

of December 21, 1997 (‘‘Merger

Agreement’’). Under the Merger