52578 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

the licensee’s presentation, followed in

turn by an opportunity for the licensee

to respond to the complainant’s

presentation. In cases where the

complainant is unable to attend in

person, arrangements will be made for

the complainant’s participation by

telephone or an opportunity given for

the complainant to submit a written

response to the licensee’s presentation.

If the licensee chooses to forego an

enforcement conference and, instead,

responds to the NRC’s findings in

writing, the complainant will be

provided the opportunity to submit

written comments on the licensee’s

response. For cases involving potential

discrimination by a contractor or vendor

to the licensee, any associated

predecisional enforcement conference

with the contractor or vendor would be

handled similarly. These arrangements

for complainant participation in the

predecisional enforcement conference

are not to be conducted or viewed in

any respect as an adjudicatory hearing.

The purpose of the complainant’s

participation is to provide information

to the NRC to assist it in its enforcement

deliberations.

* * * * *

Dated at Rockville, Maryland, this 3rd day

of October 1997.

For the Nuclear Regulatory Commission.

John C. Hoyle,

Secretary of the Commission.

[FR Doc. 97–26690 Filed 10–7–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Nuclear

Waste; Notice of Meeting

The Advisory Committee on Nuclear

Waste (ACNW) will hold its 95th

meeting on October 21, 1997, at the

William F. Bolger Center For Leadership

Development, 9600 Newbridge Drive,

Potomac, Maryland, and October 22–23,

1997, in Room T–2B3, 11545 Rockville

Pike, Rockville, Maryland.

The entire meeting will be open to

public attendance. The schedule for this

meeting is as follows:

Tuesday, October 21, 1997–8:30 a.m.

until 6:00 p.m.

Wednesday, October 22, 1997–8:30 a.m.

until 6:00 p.m.

Thursday, October 23, 1997–8:30 a.m.

until 4:00 p.m.

A. ACNW Retreat—The Committee

members will discuss their mission,

planned accomplishments, priorities,

and work processes for FY 1998–99. The

retreat will be held on October 21, 1997,

at the William F. Bolger Center For

Leadership Development.

B. Meeting with NRC’s Director,

Division of Waste Management, Office

of Nuclear Material Safety and

Safeguards—The Committee will meet

with the Director to discuss technical

assistance, developments at the Yucca

Mountain project, resources, and other

items of mutual interest.

C. Review of NRC Research and

Technical Assistance—The Committee

will review activities of NRC’s Office of

Nuclear Materials Safety and Safeguards

and Nuclear Regulatory Research in the

area of nuclear waste disposal. The

ACNW will provide input to the

Advisory Committee on Reactor

Safeguards’ February 1998 report to

Congress on NRC research.

D. Prepare for Next Meeting with the

Commission—The Committee will

prepare for its next formal meeting with

the Commission. The Committee is

scheduled to discuss items of mutual

interest with the Commission on

December 17, 1997.

E. Preparation of ACNW Reports—

The Committee will discuss planned

reports, including a recommended

approach to implement the defense-indepth concept in the revised 10 CFR

Part 60, the Application of Probabilistic

Risk Assessment Methods to

Performance Assessment in the NRC

High-Level Waste Program, ACNW

priority issues for 1998, and other topics

discussed during the meeting as the

need arises.

F. Committee Activities/Future

Agenda—The Committee will consider

topics proposed for future consideration

by the full Committee and Working

Groups. The Committee will discuss

ACNW-related activities of individual

members.

G. Miscellaneous—The Committee

will discuss miscellaneous matters

related to the conduct of Committee

activities and organizational activities

and complete discussion of matters and

specific issues that were not completed

during previous meetings, as time and

availability of information permit.

Procedures for the conduct of and

participation in ACNW meetings were

published in the Federal Register on

September 2, 1997 (62 FR 46382). In

accordance with these procedures, oral

or written statements may be presented

by members of the public, 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 Committee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the Chief, Nuclear Waste Branch, Mr.

Richard K. Major, as far in advance as

practicable so that appropriate

arrangements can be made to schedule

the necessary time during the meeting

for such statements. Use of still, motion

picture, and television cameras during

this meeting will be limited to selected

portions of the meeting as determined

by the ACNW Chairman. Information

regarding the time to be set aside for this

purpose may be obtained by contacting

the Chief, Nuclear Waste Branch, prior

to the meeting. In view of the possibility

that the schedule for ACNW meetings

may be adjusted by the Chairman as

necessary to facilitate the conduct of the

meeting, persons planning to attend

should notify Mr. Major as to their

particular needs.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting Mr. Richard K.

Major, Chief, Nuclear Waste Branch

(telephone 301/415–7366), between 8:00

A.M. and 5:00 P.M. EDT.

ACNW meeting notices, meeting

transcripts, and letter reports are now

available on FedWorld from the ‘‘NRC

MAIN MENU.’’ Direct Dial Access

number to FedWorld is (800) 303–9672;

the local direct dial number is 703–321–

3339.

Dated: October 2, 1997.

John C. Hoyle,

Acting Advisory Committee Management

Officer.

[FR Doc. 97–26692 Filed 10–7–97; 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
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52579

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 September

15, 1997, through September 26, 1997.

The last biweekly notice was published

on September 24, 1997 (62 FR 50000).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By November 7, 1997, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

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.
52580 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50-528, STN 50-529,

and STN 50-530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of amendments request: March

18, 1997, as supplemented by letters

dated July 28, 1997 and September 9,

1997

Description of amendments request:

The amendments would revise the

operating licenses for Palo Verde Units

1, 2 and 3 to reflect approval of

Amendment 42 to the Palo Verde

Nuclear Generating Station (PVNGS)

Physical Security Plan. Amendment 42

would revise the methods used to

search materials, packages and

personnel prior to their entry into the

protected area, as described within the

security plan.

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 do not significantly

increase the probability or consequences of

an accident previously evaluated. The

‘‘accident’’ as it relates to the Security Plan

would have to be an impact to the Design

Basis Threat (DBT) postulated for PVNGS.

This change does not decrease the overall

security systems (as described in paragraph’s

(b) through (h) of 10 CFR 73.55) ability to

protect PVNGS with the objective of high

assurance against the DBT of radiological

sabotage as stated in 73.1(a). This change

does not delete or contradict any regulatory

requirements.

The applicable design basis threat is

described in 10 CFR 73.1. Based on that

threat, the probability of an external

determined violent assault by stealth, or

deceptive actions, of several persons is

unaffected by the requested changes to the

search requirements. Similarly, an internal

threat of an insider, including an employee

(in any position) is no more likely to occur

as a result of the search techniques. The

probability of an attack with a four-wheel

drive land vehicle bomb is unaffected. Theft

or diversion of formula quantities of strategic

special nuclear material is a threat of removal

from the inside of the protected area, which

is not within the scope of this change that

only affects searches of material entering the

protected area.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The possibility of an accident of a new or

different kind has not been created because

the DBT (as described in the Security Plan

and 10 CFR 73.1) would not be changed as

a result of these changes. The changes

supplement regulatory requirements and

commitments already described in the

PVNGS Physical Security Plan.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes do not involve a

significant reduction in a margin of safety.

These changes to the personnel, material and

package search criteria are not specifically

considered in the basis for any margin of

safety. The DBT considers inside assistance

by a knowledgeable individual, however,

these changes would not assist this

individual in either sabotage or theft of

nuclear material.

The NRC staff has reviewed the

licensee’s analysis and, based on that

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involve no significant hazards

consideration.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072-3999

NRC Project Director: William H.

Bateman

Duke Energy Corporation, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request:

September 15, 1997

Description of amendment request:

The proposed license amendments

would revise the Technical

Specifications (TS) to:

1. Revise the reactor coolant system

heatup limitation curves in Figure 3.4-

2, which are applicable only to the first

10 effective full-power years (EFPYs).

The revised curves would be (a)

applicable to the first 15 EFPYs; (b)

include the latest radiation surveillance

capsule results; (c) remove instrument

margins by relocating them to a

licensee-controlled document, ‘‘Pressure

Temperature Limit Report;’’ and (d)

administratively delete certain

unneeded footnotes that exist in the

current figure.

2. Modify the actual surveillance

capsule identification listed in Table

4.4-5, ‘‘Reactor Vessel Material

Surveillance Program - Withdrawal

Schedule’’ (for Unit 2 only) and update

each unit—s lead factors and

withdrawal time.

3. Revise the power-operated relief

valve (PORV) setpoints in Section

3.4.9.3.a to less than or equal to 400

pounds per square inch gauge (psig) (as

left calibrated), allowable value less

than or equal to 425 psig (as found).

4. Make editorial changes to improve

consistency among various TS sections

to conform with the Westinghouse

Improved Standard Technical

Specifications, and update applicable

Code references.

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 analysis is presented below.

1. Will the changes involve a significant

increase in the probability or consequences

of an accident previously evaluated?No. No

previously evaluated accident was

considered to originate from use of the

heatup curves (change 1. above), the testing

and use of surveillance capsules (change 2.

above), the setpoint of PORVs (change 3.

above), and editorial changes to the TS. Also,

these items did not have any role in

previously analyzed accident scenarios and

thus no impact on accident consequences.

Therefore, these proposed changes will have
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52581

no impact on the consequences or

probabilities of any type of previously

evaluated accidents.

2. Will the changes create the possibility of

a new or different kind of accident from any

accident previously evaluated?

No. No actual plant equipment or

operating procedure will be affected by

the proposed changes. Hence, no new

equipment failure modes or accidents

from those previously evaluated will be

created.

3. Will the changes involve a significant

reduction in a margin of safety?

No. The margin of safety is associated with

confidence in the design and operation of the

plant. The changes to the TS do not involve

any change to plant design or operation.

Thus, the margin of safety previously

analyzed and evaluated is maintained.

On the basis of this analysis, 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

NRC Project Director: Herbert N.

Berkow

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: June 14,

1997

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) for the

Crystal River Nuclear Electric

Generating Plant Unit 3 (CR-3). The

proposed TS changes reflect the

operational limitations in mitigating

certain Small break loss-of-coolantaccident (SBLOCA) events. The licensee

also proposed changes to the associated

licensing and design bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below. The proposed changes are

addressed in three major parts: (1)

SBLOCA Mitigation, (2) Emergency

Diesel generator (EDG) upgrade and (3)

EDG Load Rejection Test and Steady

State Loads.

SBLOCA Mitigation

The licensee’s revised SBLOCA analyses

show that for certain sized breaks, a

combination of emergency core cooling

system (ECCS) flow to the reactor vessel and

emergency feedwater (EFW) flow to the once

through steam generators (OTSG) is needed

to provide for adequate core decay heat

removal. Due to load capacity limits on the

—A— EDG, the length of time that the motordriven emergency feed pump-1 (EFP-1)

would be available is limited. To ensure

adequate EFW system flow and core decay

heat removal, several actions would have to

be initiated. They include —A— EDG load

management, and EFW flow through the

turbine-driven emergency feedwater pump-2

(EFP-2) by opening the cross tie valve, flow

through both the high pressure injection

(HPI) pumps and EFP-1. The proposed TS

changes reflect the operational limitations

and other associated required actions to

ensure adequate ECCS and EFW cooling

capability remains. These changes for system

cross train dependencies and EDG load

management are required for the remainder

of current Cycle 11 only.

1. The proposed Technical Specification

changes, modifications, and operator actions

involving SBLOCA mitigation will not result

in a significant increase in the probability of

an accident previously evaluated. In

addition, the portions of the change

involving cross-train dependencies and load

management are being requested for the

remainder of Cycle 11 only, which limits the

impact on any previously established

probabilities. The initiators of any design

basis accident is not affected by the proposed

Technical Specification changes,

modifications, and operator actions involving

SBLOCA mitigation. Consequently, there is

no significant impact on any previously

evaluated accident probabilities.

The proposed Technical Specification

changes, modifications and operator actions

involving SBLOCA mitigation do not result

in a significant increase in the consequences

of SBLOCA mitigation-related accidents

previously evaluated. In this regard, the

proposed Technical Specification changes,

modifications and operator actions will not

adversely affect the integrated ability of the

EDGs and the EFW, SW [service water], RW

[raw water], Control Complex Cooling, ECCS,

DC [Decay Heat Closed Cycle Cooling Water

System], Decay Heat Seawater, and Electrical

Distribution Systems to perform their

intended safety functions. Therefore, the

combined ability of these components and

systems and actions to mitigate the

consequences of a SBLOCA will continue to

be maintained. In fact, the collective impact

of these Technical Specification changes,

modifications and operator actions represents

a restoration of the ability to mitigate the

consequences of a SBLOCA, which are

consistent with the consequences assumed in

licensing and design basis for CR-3. For

example, the installation of EFW cavitating

venturis and the improved operational range

of the turbine driven feedwater pump

increase the ability of the EFW system to

mitigate the consequences of a SBLOCA. In

addition, the Technical Specification

changes, modifications and operator actions

do not significantly affect the onsite or offsite

doses which remain a small fraction of 10

CFR Part 100 limits.

2. The proposed Technical Specification

changes, modifications and operator actions

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated. The Technical

Specification changes, modifications, and

operator actions do not involve a different

initiator for any design basis accident and do

not create new design basis scenarios.

SBLOCA mitigation, utilizing a combination

of automatic and manual actions, is already

part of the CR-3 licensing basis. Manual

operator actions necessary for the mitigation

of SBLOCAs are currently addressed or are

being addressed in EOPs [emergency

operating procedures]. Also, these Technical

Specification changes, modifications and

operator actions restore the ability to mitigate

the impact of a SBLOCA, which is consistent

with the CR-3 licensing and design basis.

Based on the above, a new or different kind

of accident does not result from this

submittal.

3. The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant reduction in the

margin of safety for SBLOCA mitigation. The

Technical Specification changes,

modifications and operator actions for the

EDGs and the EFW, SW, RW, Control

Complex Cooling Systems represent a

restoration of the overall margin of safety to

a degree that it will be consistent with the

existing plant design and licensing bases for

SBLOCA mitigation.

EDG upgrade

This aspect of the proposed license

amendment involves increases in the service

ratings of the EDGs. The required amount of

fuel oil in the EDG fuel day tank and fuel

storage tank, and lube oil storage is being

increased to ensure that adequate volume is

available to support the new service ratings.

The EDG refueling interval load test

parameters are being revised to reflect the

increased service ratings and to ensure that

the minimum test load is equal to or greater

than the expected maximum steady state

accident load. Additionally, associated EDG

Surveillance Requirements (SR) Bases are

being revised.

1. The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant increase in the

probability of an accident previously

evaluated because neither the EDGs nor the

EDG—s fuel oil and lube oil systems serve as

the initiator for any design basis accident

and, therefore, do not significantly impact

any previously evaluated accident

probabilities.

The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant increase in the

consequences of an accident previously

evaluated because the ability of the EDGs and

the EDG fuel oil and lube oil to perform their

intended safety function has not been

adversely affected. The EDGs and the EDG

fuel oil and lube oil systems remain fully

capable of performing their safety function

for all design basis accidents. The increase in

loading permitted under these changes will

reflect the manufacturer—s certified

capabilities of the EDGs. Also, the increase in

the required fuel remains within the

capabilities of the fuel tanks. The same

potential design basis failures that existed

prior to the EDG upgrades will continue to
52582 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

exist subsequent to the modifications. It

follows that the consequences of such

failures will remain a small fraction of 10

CFR Part 100 limits.

2. The proposed Technical Specification

changes, modifications and operator actions

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated. Also, the proposed

Technical Specification changes,

modifications and operator actions do not

involve any new accident initiators, or a new

or different kind of accident from any

previously evaluated. In addition, the

configuration and basic function of the EDGs

and EDG’s fuel and lube oil systems are

unaffected by the changes. In fact, the EDG

upgrades ensure that the previously

evaluated accidents are consistent with

system and component capabilities and the

current design and licensing bases.

3. The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant reduction in the

margin of safety. The EDGs and EDG—s fuel

and lube oil systems will continue to able to

be perform their safety function for all design

basis accidents. There is an increase in the

net margin of safety for fuel and lube oil

storage since required volumes have been

recalculated and increased, additional

margin has been added to the calculated

results, and the required volumes are based

on usable tank volumes instead of tank

capacity. These volumes continue to bound

the postulated worse-case accident scenario.

The increase in fuel storage required by the

changes remains within the capacity of the

storage tanks. The Technical Specification

changes, modifications and operator actions

further ensure that margins provided in

current design and licensing bases are

satisfied.

EDG Load Rejection Test and Steady State

Loads

The proposed changes for this part affects

the TS Bases. The basis of the EDG load

rejection test is being revised to bound the

largest single load. A description of ‘‘steady

state’’ is being provided with examples of

short duration loads and loads imposed by

the starting of motors. Also, addressed is the

licensee’s conclusion that the refueling

interval EDG load test is not invalidated by

loads imposed by the starting of motors.

1. The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant increase in the

probability of an accident previously

evaluated because the EDG load tests and

load rejection test do not serve as the initiator

for any design basis accident and, therefore,

do not significantly impact any previously

evaluated probabilities.

The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant increase in the

consequences of an accident previously

evaluated because the changes do not affect

the ability of the EDGs to perform their

intended safety function. Rather, the

Technical Specification changes,

modifications and operator actions provide

further assurance that the EDGs are capable

of performing their safety function. Failure of

an EDG has the same consequences as it

would if the changes were not made. It

follows that the 10 CFR Part 100

consequences of such failures has not

changed.

2. The proposed Technical Specification

changes, modifications and operator actions

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the changes do

not affect the ability of the EDGs to perform

their intended safety function. The

configuration and basic function of the EDGs,

including accurately describing the

manufacturer certified EDGs service ratings

and steady state loads, do no create a

possibility for a new or different kind of

accident. Although the load rejection test is

for an increased EDG largest single load, the

kind of accident addressed by both the load

rejection test and the refueling load test

remain the same.

3. The proposed Technical Specification

changes, modifications and operator actions

do not involve a significant reduction in the

margin of safety. The calculated loads

imposed by the starting of motors are short

duration, have a low probability of

occurrence, and are expected to be within the

manufacturer limits. In fact, the margin

confirmed by EDG refueling load testing and

load rejection testing will demonstrate a

restoration of design and licensing margin

and confirm that the EDGs remain fully

capable of performing their safety function

for all design basis accidents.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC - A5A, P. O. Box

14042, St. Petersburg, Florida 33733-

4042

NRC Project Director: Frederick J.

Hebdon

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50-336,

Millstone Nuclear Power Station, Unit

No. 2, New London County, Connecticut

Date of amendment request:

September 16, 1997

Description of amendment request:

The proposed changes to the Technical

Specifications (TSs) would modify TS

3.7.1.1, ‘‘Plant Systems Turbine Cycle

Safety Valves.’’ During its effort to verify

the current design and licensing bases

for Millstone, Unit 2, NNECO has

determined that the maximum

allowable power level high trip

setpoints with inoperable steam line

code safety valves specified in Table

3.7-1 of TS 3.7.1.1 are incorrect. The trip

setpoints were not changed to be

consistent with a previously approved

reduction in the maximum power level

high trip setpoint. In addition, NNECO

is also in the process of reanalyzing the

inadvertent closure of the main steam

isolation valve (MSIV) and the loss of

electrical load events. The results of the

reanalysis indicate that the MSIV event

results in the highest peak pressure in

the secondary system and that the

formula currently contained in the TS

Bases for TS 3.7.1.1 may not result in

the correct trip setpoints.

Specifically, NNECO proposes to: (1)

delete TS Table 3.7.1 by not allowing

operation in Mode 1 or 2 with

inoperable steam line code safety

valves, (2) modify the associated action

statement in TS 3.7.1.1, and (3) update

the TS Bases to reflect the proposed

changes and update the amendment

history numbers to reflect previously

approved amendments.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change does not involve an

SHC [significant hazards consideration]

because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This proposed change will remove the

ability to operate in Modes 1 or 2 with

inoperable main steam line code safety

valves. Operation in Mode 3 will be retained,

provided no more than three main steam line

code safety valves per steam generator are

inoperable.

The primary function of the main steam

line code safety valves is to prevent

secondary system overpressurization. These

valves will also provide reactor core heat

removal and design basis accident mitigation.

This proposed change does not affect the

length of time the plant can operate with

inoperable main steam line code safety

valves before compensatory actions must be

taken. (Four hours is still allowed to restore

the valve(s) to operable status.) This

proposed change does not affect the

probability of occurrence of any design basis

accident and does not affect how the main

steam line code safety valves function to

mitigate design basis accidents. Therefore,

this change does not significantly increase

the probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not alter the

way any structure, system, or component

functions. The proposed change will

conservatively change plant operation in

Modes 1 and 2 by removing the ability to

operate at power with inoperable main steam
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52583

line code safety valves as currently specified

in Technical Specification 3.7.1.1. It does not

introduce any new failure modes and does

not alter any assumption made in the safety

analysis.

Therefore, the change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

This proposed change to Technical

Specification 3.7.1.1 will remove the ability

to operate in Modes 1 or 2 with inoperable

main steam line code safety valves.

Operation in Mode 3 will be retained,

provided no more than three main steam line

code safety valves per steam generator are

inoperable. The operability of the main steam

line code safety valves ensures that the

secondary system pressure will be limited to

within 110% (1100 psig) of the design

pressure of 1000 psig during the most severe

anticipated system operational transient.

This change will not affect the operability

requirements for the main steam line code

safety valves and will not affect the length of

time the plant can operate with inoperable

main steam line code safety valves before

compensatory actions must be taken. This

will ensure the plant equipment required for

design basis accident mitigation will be

available. Therefore, there is no significant

reduction in a margin of safety as defined in

the Bases of Technical Specification 3.7.1.1.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut

NRC Deputy Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: August

29, 1997

Description of amendment request:

Based on a review and subsequent

calculations of the cold

overpressurization protection (COPS)

enabling temperature and the

emergency core cooling system (ECCS)/

charging system Mode 3 requirements,

NNECO proposes to reduce the COPS

enabling temperature. As a result,

NNECO proposed the following

Technical Specifications (TS) changes:

new heatup and cooldown pressure/

temperature limit curves and their

associated requirements; new power

operated relief valve (PORV) setpoint

curves and their associated

requirements; revisions to the reactor

coolant loops and coolant circulation,

ECCS, boration systems, and COPS to

incorporate the lower enabling

temperature and new restrictions for

cold overpressure protection system

(COPPS), PORV undershoot, and

residual heat removal (RHR) relief valve

bellows; addition of a footnote to allow

a reactor coolant pump (RCP) to

substitute for an RHR pump during

heatup from Mode 5 to Mode 4, which

is consistent with the improved

standard technical specification (STS);

reword TS 3/4.4.9.3 and its Bases

section to be consistent with the

improved STS; and revision of the

affected Bases sections to be consistent

with the proposed changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

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

satisfied. The proposed revision does not

involve [an] SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

Probability of Occurrence of Previously

Evaluated Accidents

Since the PORV setpoints and the COPS

enabling temperature have been calculated in

accordance with 10CFR50, Appendix G and

ASME [American Society of Mechanical

Engineers] Section XI, the change will not

alter the probability that an

overpressurization event will result in a loss

of RV [reactor vessel] integrity. The new

PORV setpoint curves are lower than the

current curves in certain temperature ranges

(below approximately 130—F and above

approximately 220°F), and therefore the

operating window is slightly decreased.

However, the reduced operating window is

still sufficient for normal anticipated

pressure fluctuations. Below 160°F, operation

of Reactor Coolant Pumps are prohibited if

the PORVs are armed for COPPS; therefore,

PORV actuation will not occur below 160—F

when the RCPs are running. In a water solid

condition, RCS [reactor coolant system]

pressure is maintained via the letdown low

pressure control valve, which, when in

automatic mode, maintains the RCS pressure

in a relatively narrow range. When the RCPs

are not running, the PORV COPPS system

can be actuated. However, for this condition,

the allowable pressure range is 0 to 418 psia

[pounds per square inch atmospheric]. This

pressure range is sufficient to accommodate

normal anticipated pressure fluctuations.

Above 220°F, the minimum pressure range

is from 300 psia to 595 psia; this range is

sufficient to accommodate normal

anticipated pressure fluctuations. In this

temperature range, a pressurizer bubble is

normally present, which will minimize any

pressure fluctuations, thereby limiting the

possibility of a PORV actuation. Based on

this, it is concluded that the proposed change

will not impact the probability of occurrence

that a PORV will be challenged.

When the RHR relief valves are used for

COPS there is no credible scenario which

would result in excessive relief valve

undershoot. This is because these valves are

spring loaded relief valves which are

designed to close whenever the RCS pressure

decreases below the nominal setpoint of 440

psig [pounds per square inch gauge]. This

provides assurance that there will be no

damage to the seal of a running RCP.

The proposed changes to the heatup/

cooldown curves and the reduction in the

enabling temperature for COPS only affect

operational limits and can not be initiators of

an event. The restrictions on RC [reactor

coolant], RHR and ECCS pump operation can

not result in an event initiator. Two separate

operator actions are required to start an ECCS

or RC pump. These two necessary actions as

well as procedural controls are sufficient to

prevent an inadvertent ECCS or RC pump

start. De-energizing the RCPs when returning

a loop to service can not initiate an event.

The proposed change will provide an

operable charging pump to ensure RCP seal

flow and reactivity control will be available.

When the RCP is in operation, the charging

pump provides the preferred method for seal

flow. The proposed change minimizes the

time that this preferred method is

interrupted. A loss of charging pump seal

flow will not cause a malfunction of an RCP

because the pump is designed to use RCS

flow as an alternate method at these

conditions. Not allowing two charging

pumps to run simultaneously and requiring

at least one pump to be in pull-to-lock,

assures a second pump will not start on an

inadvertent SI [safety injection] and exceed

the assumptions in the Appendix G analysis

or initiate a Boron Dilution or CVCS

[chemical and volume control system]

Malfunction event. If an operator were to

inadvertently start the second pump, a failure

of the charging throttle valve, FCV-121, and

one relief valve credited for COPS would be

necessary to exceed the assumptions in the

Appendix G analysis. In addition, the actual

time allowed for swapping the charging

pumps is short. The remainder of the hour

allows for documented verification of the

disabling of the required pump. The

proposed change will not change any control

systems for these pumps or alter the system

configuration that would affect the

probability of an uncontrolled increase in

charging flow. The procedure requirements

to swap pumps and the likelihood of these
52584 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

multiple failures occurring during the short

duration allowed in this footnote provide

adequate assurance that an

overpressurization event will not occur.

Maintaining at least one pump always

operable makes the system more reliable for

reactivity control than the current method

which disables both pumps simultaneously.

The proposed change to maintain one

charging pump operable in Mode 4 [cannot]

initiate an event because of the stable

reactivity condition of the reactor, the

emergency power supply requirement for the

operable charging pump, and the fact that the

plant is procedurally required to be borated

to the highest required boron concentration

for Modes 3, 4, or 5 prior to entering Mode

4. These changes do not effectively change

the availability of plant equipment or the

way that the plant is operated.

The proposed change to substitute an RCS

loop for an RHR loop during a planned

heatup, can not initiate an event. The RCP

will be verified as operating properly prior to

stopping the RHR pump and as such will not

initiate a loss of decay heat removal (by

heating up to steam the SGs [steam

generators])/loss of flow. While the RCP is in

operation, it performs the RHR boron mixing

function and the decay heat removal function

is not required for heatup. Using the RCP to

perform this function will not affect the

probability that the RCP could fail because it

will be operated within its normal operating

design conditions. Aligning RHR in the ECCS

lineup will not affect the probability of a

RHR pump to start. The pump will be

operable in this lineup. Currently in Mode 5,

RHR is lost on a LOP [loss of offsite power]

and is manually restarted once the diesel is

running. With the proposed change, the RCP

will be lost on a LOP and the RHR pump will

have to be manually started. Thus, the

proposed change does not affect the

probability that the RHR pump could fail.

Since the current response to a LOP is to

manually restart the RHR pump, operator

action is needed independent of this change.

The proposed change allows normally open

valves to be closed in Mode 5 to align RHR

for ECCS injection. This introduces

additional manual actions which could

extend the time required to establish flow. In

addition, if one diesel generator were to fail,

manual operation of a valve in the ESF

[engineered safety features] building would

be necessary. The mechanistic ’failure to

open’ of valves that is introduced by the

change as well as the need for manual

operator action to realign these valves

increases the time to establish heat removal.

However, there is sufficient time to reestablish RHR because this note applies only

for a heatup in which the plant will have

been shutdown for at least several hours

which causes decay heat to be low (as

compared to high decay heat immediately

following a plant trip). Thus, it is concluded

that there is no impact on the probability of

failure of RHR to perform its required

function.

The proposed change to the ECCS wording

does not result in any new failure modes that

could initiate an event since manual

realignment from the control room is

currently allowed. Nor can the manual

alignment of RHR valves initiate an event

because this alignment is only for accident

mitigation.

Therefore, the proposed changes do not

increase the probability of occurrence of

previously evaluated accidents.

Consequences of Previously Evaluated

Accidents

The revised Pressure/Temperature curves

were calculated in accordance with 10CFR50,

Appendix G, ASME Section XI, and

Regulatory Guide 1.99, Revision 2. This

provides assurance that an inadvertent

overpressurization event will not result in a

loss of RV integrity. The restrictions on RCP

operation and the requirement to de-energize

the RCPs in Modes 5 and 6 when returning

a loop to service are consistent with the

assumptions made in this Appendix G

analysis and the RCPs are not required for

accident mitigation for any previously

evaluated accidents and therefore do not

affect the consequences.

The COPS relieving capability is greater

than the maximum RCS pressurization rate

resulting from any allowed pump

combinations, and the PORV setpoints have

been adjusted to take into account

instrumentation effects. This will provide

assurance that COPS will continue to

perform its safety function. Since the COPS

enabling temperature has been demonstrated

to be conservative at 275—F, allowing SI

pump operability above 275—F will have no

impact on vessel non-ductile failure.

The restriction between 275—F and 350—F

on the SI and charging pumps, has been

appropriately moved to the reactor coolant

loop section to provide protection for the

RHR system (RCS protective boundary) and

to the cold overpressure protection section to

provide protection for the RHR relief valves

and the RCP seals. By incorporating this

requirement previously located in the ECCS

TS, RCS integrity is ensured.

With the RCS less than 160—F, the

consequences of the PORV undershoot from

the proposed PORV setpoints are that the

RCS pressures may drop below the minimum

requirement for RCP seal integrity. However,

no seal damage will occur since a

requirement has been added prohibiting the

operation of RCPs below 160°F with the

PORVs not isolated while in the low setpoint

mode. With cold overpressure relief valves in

service above the COPS enable temperature

(275°F), restrictions are placed on the startup

of an RCP and the number of ECCS pumps

capable of injecting into the RCS to prevent

unacceptable mass or energy addition

transients. This provides assurance that the

RHR relief valve capacity will not be

exceeded and that PORV undershoot will not

challenge the RCP ι1 seal. The restriction on

the maximum number of ECCS pumps

ensures that the integrity of the RHR relief

valve bellows and the RCP seals during mass

injection transients (i.e., inadvertent SI).

The restrictions on RCS/SG secondary side

temperature mismatch ensure that an

unanalyzed energy addition event does not

occur when an RCS loop is placed in

operation.

The consequences of a small break LOCA

[loss of coolant accident] in COPS Mode 4 are

not affected because the plant will continue

to maintain one charging pump operable in

Mode 4. In addition, additional options are

provided in the bases of TS 3/4.4.9.3 for

disabling the required charging and SI pumps

that will allow faster restoration if required

to mitigate a LOCA or loss of RHR in Modes

4, 5 and 6.

An RHR pump will remain available in

Mode 4 with manual realignment from the

control room as required to perform its ECCS

safety function. The changes have no impact

on the capability of RHR to function in the

ECCS mode. RHR is credited during a safety

grade cold shutdown. The proposed change

assures that the RHR system will be available

to perform its heat removal function during

a safety grade cold shutdown and thus, there

is no change in the analysis assumptions or

consequences.

The changes also eliminate an

inconsistency between the charging system

operability requirements for boration and the

charging system operability requirements for

cold overpressure protection. The

requirement to maintain two charging pumps

operable in Mode 4 will be reduced to one

charging pump. As stated in the proposed

basis section, a second method of boration is

not required to be OPERABLE in Mode 4 for

single failure considerations based on the

stable reactivity condition of the reactor, the

emergency power supply requirement for the

operable charging pump, and the fact that the

plant is procedurally required to be borated

to the highest required boron concentration

for Modes 3, 4, or 5 prior to entering Mode

4. This provides assurance that reactivity

control will be maintained and stable while

only one charging pump is operable for cold

overpressure concerns. These changes do not

effectively change the availability of plant

equipment or the way that the plant is

operated. The changes will not adversely

impact the assumption for the limiting

dilution flow path and flow rate and

therefore, the consequences of a boron

dilution event are not affected.

The proposed changes will maintain a

charging pump operable for reactivity control

while ensuring that the flow limits in the

Appendix G analyses are not exceeded.

Remaining within the bounds of the

Appendix G limits ensures reactor vessel

integrity in Mode 4. Since the change

maintains the reactor vessel integrity, it does

not introduce any means of releasing

radionuclides post-accident. The

consequences of a small break LOCA in

Mode 4 are not affected because the plant

will continue to maintain one charging pump

operable in Mode 4. These changes are

reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and

3.1.2.4. Adequate protection is provided for

reactor vessel integrity while maintaining

reactivity control operability.

In Mode 5, RHR requirements are specified

for decay heat removal in the case of a loss

of offsite power but none are specified for

ECCS accident mitigation. The first RHR train

will be aligned for injection prior to taking

the second train out of service. This provides

assurance that this train will be available if

needed in Mode 5. Currently in Mode 5,

following a LOP the RHR system can be reestablished by restarting the RHR pump once

the diesel is running. No valve manipulations
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52585

are necessary. With the proposed change,

when the operating RCP trips following a

LOP, some of the RHR valves must be

realigned from the ECCS to heat removal

mode. If one diesel generator were to fail,

manual operation of a valve in the ESF

building would be necessary. Since this

footnote is only applicable during a heatup,

decay heat will be low. There is sufficient

time to re-establish RHR even if action

outside the control room is necessary. Since

there are four operable RCS loops, a bubble

drawn in the pressurizer and the RCS

pressurized, the plant will heat up to Mode

4 and natural circulation will provide core

cooling if the RHR system cannot be reestablished. Thus, decay heat removal is

assured and there is no affect on the

consequences of a LOP.

Since the structural integrity of the RCS is

maintained and adequate core cooling and

reactivity control will be available for design

basis events, the proposed changes will have

no adverse impact on the consequences of

previously evaluated accidents.

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 temperature/pressure limits will

continue to meet the requirements of

10CFR50, Appendix G. Since the new limits

continue to provide assurance of reactor

vessel integrity, the proposed change does

not create the possibility of an accident of a

different type than previously evaluated.

Adequate RCS pressure-relieving capabilities

will continue to be maintained throughout

the shutdown modes. No new malfunctions

will be introduced which could result in a

new accident postulated in Modes 3-5.

The restrictions on RCP operation do not

create the potential for unanalyzed heat

injection transient as a result of an

inadvertent RCP start because two operator

actions are required to start a pump. The

requirement to have all RCPs de-energized,

prior to unisolating a loop adds additional

assurance that an energy addition transient

will not occur.

The proposed change to allow 2 charging

pumps to be operable does not create an

accident of a different type because there will

be adequate controls to ensure that the

second pump does not inadvertently start

and initiate an increase in RCS inventory or

a boron dilution. Procedural controls will

minimize the amount of time that both

charging pumps are operable and at no time

will two pumps be out of pull-to-lock.

The proposed footnote to TS 3.4.1.4.1 to

remove RHR heat removal from operation

allows normally open valves to be closed in

Mode 5 to align RHR for ECCS injection. This

introduces ’failure to open’ as a potential

mechanistic failure malfunction in the RHR

system. This is a malfunction of a different

type since previously stroking of these valves

was not needed to establish RHR. The current

response to a LOP is to manually restart the

RHR pump only, with no valve

manipulations required. The proposed

change adds the manual action of realigning

the valves. Since operator action to reestablish RHR following a LOP is required

independent of the proposed changes,

crediting operator action does not create the

potential for a malfunction of a different

type. Allowing both trains of RHR to be out

of service does not create a different accident

because additional requirements have been

specified for RCS loop operability and at

least one RHR pump is operable for ECCS

when the core cooling requirement is being

met by crediting RCS loop operability.

Meeting the Mode 4 TS conditions prior to

heatup, ensures two diesels are operable. As

such, a single failure would only require one

valve to be manually realigned in the ESF

building. Adequate time is available to

accomplish these actions since this note only

applies during heatup, when decay heat is

very low. Further, with four RCS loops

operable and a bubble drawn in the

pressurizer and the RCS pressurized, the

steam generators can be used for core cooling

via natural circulation once the plant heats

up to Mode 4, in the event the RHR cannot

be re-established. Since core cooling will be

assured if a LOP occurred during heatup in

Mode 5, the change in plant response to this

event does not constitute an accident of a

different type.

The proposed changes to TS 3.5.3.f to

manually realign the ECCS valves is no

different from what is currently evaluated.

During a Mode 4 LOCA adequate procedural

guidance is provided to ensure that RHR will

be realigned for injection. The proposed

change allows RHR to be aligned to perform

its safety grade cold shutdown heat removal

function.

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 new proposed curves raises the lower

bound on RCS temperature, resulting in

increased RCS ductility and therefore

increased structural margin against nonductile failure. The new curves take into

account the dynamic pressure effects

identified in NRC Information Notice 93-58

and are calculated in accordance with

10CFR50 Appendix G, ASME Section XI and

Regulatory Guide 1.99, Revision 2. These

changes to the P/T [pressure/temperature]

limits are reflected in TS 3.4.9.1. Additional

restrictions have been placed on RCP

operation to ensure that assumptions used in

developing the curves remain bounding.

These are also reflected in TS 3.4.1.3,

3.4.1.4.1, 3.4.1.4.2 and 3.4.1.6. As such, the

curves will continue to provide the required

assurance for reactor vessel integrity.

The COPS enable temperature is proposed

to be lowered from the current 350°F to

275°F which provides a margin of 31—F

above that required by NRC Branch

Technical Position RSB 5-2. The reduction of

the COPS enabling temperature eliminates

the need for COPS to be operable in Mode

3. This will simplify the transition between

Mode 3 and Mode 4.

Additional changes have been made to the

Overpressure Protection TS to ensure that the

assumptions made in the Appendix G

calculations remain bounding. These include

additional restrictions on charging pump and

SI pump operability and the modification of

the PORV setpoints. The pump requirements

have been transferred from the ECCS

specification and expanded to cover Modes

4, 5 and 6. In addition, these same pump

restrictions have been included in TS 3.4.1.3

whenever RHR is in service. This provides

added assurance that the RHR piping will not

be overpressurized by an inadvertent

actuation of an SI or charging pump.

Additional actions and surveillances have

been provided to assure that assumptions on

charging pump and SI pump operability will

be met. The additional options for assuring

the inoperability of the SI and charging

pumps require two distinct operator actions

to restore injection capability from these

pumps. Thus, these options are equivalent in

providing assurance that an inadvertent

injection will not occur while at the same

time allowing faster restoration if needed to

mitigate a loss of RHR.

A requirement to have all RCPs deenergized, prior to unisolating a loop is

added to TS 3.4.1.6.c, to ensure that loop

flow will not be initiated which results in an

energy addition transient from the secondary

side of the SG being unisolated. This change

will preclude RCS overpressurization when

an idled loop is returned to service and SG

secondary side temperature is greater than

the RCS temperature.

The PORV setpoints were established to

ensure that the P/T limit curves are not

exceeded as a result of a single operator

action or as a result of a single equipment

malfunction, as required by the current

system design basis criteria (i.e., SRP

[standard review plan] Branch Technical

Position RSB 5-2).

A clarification of the hydrostatic and leak

test requirements ensures a uniform reactor

vessel temperature for the test. A 72 hour

time limit is placed on the performance of

engineering evaluations of out of

specification condition. This provides added

assurance for RPV [reactor pressure vessel]

integrity.

The changes also eliminate an

inconsistency between the charging system

operability requirements for boration and the

charging system operability requirements for

cold overpressure protection. These are

reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and

3.1.2.4. The Bases requirement to maintain

two charging pumps operable in Mode 4 will

be reduced to one charging pump. As stated

in the proposed basis section, a second

method of boration is not required to be

OPERABLE in Mode 4 for single failure

considerations based on the stable reactivity

condition of the reactor, the emergency

power supply requirement for the operable

charging pump, and the fact that the plant is

procedurally required to be borated to the

highest required boron concentration for

Modes 3, 4, or 5 prior to entering Mode 4.

This provides assurance that reactivity

control will be maintained and stable while

only one charging pump is available. The

additional options for disabling the charging

pump (provided in the bases for TS 4.4.9.3.5)

will allow for faster restoration when needed

while maintaining two distinct operator
52586 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

actions to prevent a second pump from being

started. This provides added assurance that

reactor vessel integrity will be maintained.

Procedures will minimize the amount of

time that both charging pumps are operable

and having at least one pump in pull-to-lock

will ensure that the second pump does not

inadvertently start and exceed the Appendix

G analysis limits and thus, ensure reactor

vessel integrity.

The TS bases for requiring RHR in Mode

5 is to remove decay heat and provide RCS

circulation. Since the RCP can perform the

RHR circulation function and the decay heat

removal function is not required during

heatup, the proposed change is consistent

with the bases. Since this option is only

allowed during heatup where decay heat is

low, sufficient time will be available to reestablish RHR heat removal as required to

mitigate a LOP in Mode 5. Further, with the

RCS pressurized, four RCS loops operable

and the SG filled, core cooling can be

accomplished by the steam generators via

natural circulation once the plant heats up to

Mode 4, in the event that RHR cannot be reestablished. Therefore, the design basis

analyses remain limiting and the margin of

safety is not reduced.

The original plant design allows the RHR

pumps to be available for both heat removal

while shutdown and ECCS. As such, an

allowance, TS 3.5.3.f, was provided to allow

manual realignment from heat removal to

ECCS mode. The specific wording of TS

3.5.3.f implies that this realignment only

involves the suction valves. Since discharge

valves must also be realigned, the TS is being

reworded to apply for the discharge as well

as suction valves. Therefore, this change is a

clarification of the existing TS.

The proposed changes do not impact the

protective boundaries (reactor vessel

integrity) nor any of the design basis

accidents.

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, CT 06141-0270

NRC Deputy Director: Phillip F.

McKee

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: July 25,

1997

Description of amendment request:

The proposed amendment request

would implement 10 CFR Part 50

Appendix J, Option B by revising the

Technical Specifications (TS) to allow

the frequency of conducting integrated

leak rate testing (ILRT) and local leak

rate testing (Type B and C) to be based

on component performance.

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 implements Option B

of 10 CFR Part 50 Appendix J on

performance-based containment leakage

testing. The proposed change does not

involve a change to the plant design or

operation. As a result, the proposed change

does not affect any parameters or conditions

that contribute to the initiation of any

accidents previously evaluated. The

proposed change potentially affects the leaktight integrity of the containment structure

designed to mitigate the consequences of a

Loss-of-Coolant Accident (LOCA). The

function of the containment is to maintain

functional integrity during and following the

peak transient pressures and temperatures

and limit fission product leakage following

the design basis LOCA. Because the proposed

change does not alter the plant design, only

the frequency of measuring Type A, B, and

C leakage, the proposed change does not

directly result in an increase in containment

leakage.

Test intervals will be established based on

the performance history of components being

tested. The frequency of monitoring the

relatively few containment isolation valves

and/or containment penetrations subject to

above normal leakage will not decrease by

implementing Option B of Appendix J. A

performance based program will identify

those valves and penetrations which must

continue to be tested each refueling outage.

The risk resulting from the proposed

changes is characterized as follows, based

primarily on the results contained in

NUREG-1493 ‘‘Performance-Based

Containment Leakage Test Program,’’ the

principal Technical Support Document used

by the NRC as the basis for the Appendix J

Final Rule:

Type A Testing

NUREG-1493 found that the effect of

containment leakage on overall accident risk

is minimal since risk is dominated by

accident sequences that result in failure or

bypass of the containment. Industry wide,

Integrated Leak Rate Tests (ILRTs) have only

found a small fraction of the leaks that

exceed current acceptance criteria. Only

three percent of all leaks are detectable only

by ILRTs, and therefore, by extending the

Type A testing intervals, only three percent

of all leaks have a potential for remaining

undetected for longer periods of time. In

addition, when leakage has been detected by

ILRTs, the leakage rate has been only

marginally above existing requirements. The

Fort Calhoun Station Unit No. 1 Type A

testing confirms the industry-wide

experience that a majority of the leakage

experienced during Type A testing is through

components tested by Type B and C tests.

NUREG-1493 found that these

observations, together with the insensitivity

of reactor accident risk to the containment

leakage rate, show that increasing the Type

A leakage test intervals would have a

minimal impact on public risk.

Type B and C Testing

NUREG-1493 found that while Type B and

C tests can identify the vast majority (greater

than 95 percent) of all potential leakage

paths, performance-based alternatives to

current local leakage-testing requirements are

feasible without significant risk impacts. The

risk model used in NUREG-1493 suggests

that the number of components tested would

be reduced by about 60 percent with less

than a three-fold increase in the incremental

risk due to containment leakage. Since, under

existing requirements, leakage contributes

less than 0.1 percent of overall accident risk,

the overall impact is very small. In addition,

the NRC’s Final Regulatory Impact Analysis

concluded that while the extended testing

intervals for Type B and C tests led to minor

increases in potential offsite dose

consequences, the beneficial expected

decrease in onsite worker dose received

during ILRT and local leak rate testing

exceeds (by at least an order of magnitude)

the potential off-site dose consequences.

Therefore, the proposed change will not

result in a significant increase in 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.

There will be no physical alterations to the

plant configuration, changes to setpoint

values, or changes to the implementation of

setpoints or limits as a result of this proposed

change. As a result, the proposed change

does not affect any of the parameters or

conditions that could contribute to initiation

of any accidents.

This change involves the reduction of Type

A, B, and C test frequency. Except for the

method of defining the test frequency, the

methods for performing the actual tests are

not changed. No new accident modes are

created by extending the testing intervals. No

safety-related equipment or safety functions

are altered as a result of this change.

Extending the test frequency has no influence

on, nor does it contribute to, the possibility

of a new or different kind of accident or

malfunction from those previously analyzed.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52587

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change only affects the

frequency of Type A, B, and C testing. Except

for the method of defining the test frequency,

the methods for performing the actual tests

are not changed.

The frequency of monitoring the relatively

few containment isolation valves and/or

containment penetrations subject to above

normal leakage will not decrease by

implementing Option B of Appendix J. A

performance based program will identify

those valves and penetrations which must

continue to be tested each refueling outage.

NUREG-1493 has determined that, under

several different accident scenarios, the

increased risk of radioactivity

release from containment is negligible with

the implementation of these proposed

changes.

The margin of safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to containment leakage rate. The

containment isolation system is designed to

limit leakage to La, which is stated in the

Fort Calhoun Station Unit No. 1 Technical

Specifications to be 0.1 percent by weight of

the containment air per 24 hours at 60 psig.

The limitation on containment leakage rate

is designed to ensure that total leakage

volume will not exceed the value assumed in

the accident analyses at the peak accident

pressure. The margin to safety for the offsite

dose consequences of postulated accidents

directly related to the containment leakage

rate is maintained by meeting the 1.0 La

acceptance criteria. The La value is not being

modified by this proposed change.

Except for the method of defining the test

frequency, no change in the method of testing

is being proposed. The Type B and C tests

will continue to be done at 60 psig or greater.

Other programs are in place to ensure that

proper maintenance and repairs are

performed during the service life of the

primary containment and systems and

components penetrating the primary

containment.

Therefore, the proposed change will not

result in a significant reduction in a margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

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

3502

NRC Project Director: William H.

Bateman

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: August

26, 1997

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

4.6.5.3.1b, for the Filtration,

Recirculation and Ventilation System

(FRVS), Ventilation Subsystem, and TS

4.6.5.3.2b for the FRVS Recirculation

Subsystem. The revised TSs would state

that the heaters should be ‘‘operating

(automatic heater modulation to

maintain relative humidity)’’ instead of

‘‘on’’ when performing the 10-hour,

monthly test.

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 TS revisions involve no

hardware changes and no changes to existing

structures, systems or components.

Conducting TS Surveillance Requirements

4.6.5.3.1.b and 4.6.5.3.2.b with the FRVS

recirculation unit and ventilation unit

heaters in automatic modulation to maintain

the relative humidity within the design

requirements, meets the intent of the USNRC

Regulatory Guide 1.52, position C.4.d, in

reducing adsorber and HEPA filter moisture

levels. In the unlikely event that the adsorber

and HEPA filters, that are enclosed and

isolated in a confined space should reach an

equilibrium at the maximum design

operating humidity level, the 10 hour run

with heaters energized would reduce the

humidity to acceptable levels. Therefore, the

proposed changes do not change the postaccident performance characteristics of the

FRVS adsorber or HEPA filters below the

design requirements and does not increase

the consequences of accidents previously

identified. Since there are no changes to the

operation of FRVS in normal or post-accident

operating conditions, there is no increase in

the probability of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes contained in this

submittal will not adversely impact the

operation of any safety related component or

equipment. PSE&G has concluded that [the]

method of performing the monthly FRVS

recirculation unit and ventilation unit

surveillances with the heaters modulating

adequately maintains and demonstrates

operability of FRVS. Since the proposed

changes involve: 1) no hardware changes; 2)

no changes to FRVS operation in normal

operating or post-accident conditions; and 3)

no changes to existing structures, systems or

components, there can be no impact on the

potential occurrence of any accident.

Furthermore, there is no change in plant

testing proposed in this change request

which could initiate an event. Therefore,

these changes will not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The revisions to TS Surveillance

Requirements 4.6.5.3.1.b and 4.6.5.3.2.b

provide a more accurately defined basis for

performing this surveillance test. The

proposed changes reflect PSE&G’s position

on satisfying USNRC Regulatory Guide 1.52,

position C.4.d. Since PSE&G has concluded

that performing TS Surveillance

Requirements 4.6.5.3.1.b and 4.6.5.3.2.b with

the FRVS recirculation unit and ventilation

unit heaters in automatic moduation [sic]

[modulation] to maintain the relative

humidity within the design requirements,

adequately reduces adsorber and HEPA filter

moisture levels, the proposed changes do not

significantly reduce a margin of safety in

FRVS. Since the FRVS recirculation units

and ventilation units will continue to be

tested with the heaters: 1) operable; and 2)

set at the demand necessary to ‘‘reduce the

buildup of moisture,’’ PSE&G believes that

the proposed changes to clarify the TS are

justified.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit - N21,

P.O. Box 236, Hancocks Bridge, NJ

08038

NRC Project Director: John F. Stolz

Rochester Gas and Electric

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

Ginna Nuclear Power Plant, Wayne

County, New York

Date of amendment request: August

19, 1997

Description of amendment request:

The proposed amendment would revise

the Ginna Station Improved Technical

Specifications (ITS) by revising the

Emergency Core Cooling System

Accumulators Surveillance Requirement

3.5.1.2 to correct the specified

accumulator borated water volume

values in order to match the associated

accumulator percent level values.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
52588 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

issue of no significant hazards

consideration, which is presented

below:

1. Operation of Ginna Station in

accordance with the proposed changes does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. The change is only to

correct a conversion error with respect to

accumulator borated water volume. This does

not increase the probability of any accident

previously evaluated since the accumulator

water volume provides mitigation capability

only (i.e., does not initiate any accident). The

affected accident analyses with respect to the

accumulator (e.g., small and large [loss-ofcoolant] LOCA and steam line break) have

been re-evaluated using the correct

accumulator water volume values with

acceptable results. Therefore, these changes

do not involve a significant increase in the

probability or consequences of an accident

previously analyzed.

2. Operation of Ginna Station in

accordance with the proposed changes does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed changes

do not involve a physical alteration of the

plant (i.e., no new or different type of

equipment will be installed) or changes in

the methods governing normal plant

operation. Ginna Station operators verify

accumulator water volume via percent level

(versus cubic feet) which remains

unchanged. Thus, this change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Operation of Ginna Station in

accordance with the proposed changes does

not involve a significant reduction in a

margin of safety. The proposed changes only

correct a conversion error. The error has been

re-evaluated with acceptable results. As such,

no question of safety is involved, and the

change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610

Attorney for licensee: Nicholas S.

Reynolds, Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005

NRC Project Director: Alexander W.

Dromerick, Acting Director

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:

February 14, 1997, as supplemented by

letters dated June 20, August 5, and

September 22, 1997

Description of amendments request:

The proposed amendments would

change the maximum reactor core

power level for facility operation from

2652 megawatts thermal (MWt) to 2775

MWt in the Farley, Units 1 and 2,

Facility Operating Licenses. In addition,

the proposed amendments would

involve the following Technical

Specification (TS) changes.

The defined rated thermal power for

Farley; departure from nucleate boiling

(DNB) parameters for reactor coolant

system (RCS) average temperature (Tavg);

pressurizer pressure; and RCS flow

would be changed.

The reactor trip system interlock

setpoint for power range neutron flux

(P-8) and engineered safety features

(ESF) actuation trip setpoint for steam

generator water high-high level for

turbine trip and feedwater isolation (P14), and ESF actuation system interlock

for low-low Tavg (P-12) would be

modified to reflect analytical results.

An evaluation of additional reactor

trip system and ESF actuation system

safety analysis limits and trip setpoints

would result in changes to the allowable

values for several functions.

On the basis of the results of new

containment analyses, the maximum

peak calculated containment internal

pressure for a loss-of-coolant accident

(LOCA) event would be revised. The

main steamline isolation valve closure

time requirement would be revised.

Surveillance requirements for

emergency core cooling systems (ECCS)

would be modified to reflect reduced

ECCS flows. The number of secondary

system hydrostatic pressure tests (Table

5.7-1) would be increased. For Farley

Unit 2 only, the steam generator F*

distance would be revised.

Changes to the plant design features

and administrative controls are also

proposed. These changes would revise

the RCS fluid volume contained in

Section 5.4 and the addition of the NRCapproved references for best estimate

LOCA listed in Section 6.9.1.

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:

DEFINITION, DESIGN FEATURE AND

ADMINISTRATIVE CONTROL CHANGES

* * * *

1. The proposed changes to the rated

thermal power definition, RCS fluid volume,

and COLR [Core Operating Limit Report]

references do not increase the probability or

consequences of an accident previously

evaluated in the FSAR [Final Safety Analysis

Report]. The comprehensive analytical efforts

performed to support the proposed uprating

included a review and evaluation of all

components and systems (including interface

systems and control systems) that could be

affected by this change. The revised power

uprate value and RCS fluid volume were

inputs to applicable safety analyses. All

systems will function as designed, and all

performance requirements for these systems

have been evaluated and found acceptable.

None of these proposed changes directly

initiate any accident; therefore, the

probability of an accident has not increased.

All dose consequences have been analyzed or

evaluated with respect to these parameters,

and all acceptance criteria continue to be

met. Therefore, the consequences of an

accident previously evaluated in the FSAR

have not increased.

2. The proposed changes do not create the

possibility of a new or different kind of

accident than any accident already evaluated

in the FSAR. No new accident scenarios,

failure mechanisms or limiting single failures

are introduced as a result of the proposed

changes. The proposed technical

specification changes have no adverse effects

on any safety-related system and do not

challenge the performance or integrity of any

safety-related system. Therefore, the

possibility of a new or different kind of

accident is not created.

3. The proposed operating license and

technical specification changes do not

involve a significant reduction in a margin of

safety. All analyses supporting the proposed

power uprate reflect the RCS fluid volume

and rated thermal power values. The use of

NRC approved BELOCA [best estimate

LOCA] methodology must be referenced

since BELOCA will now be the LBLOCA

[large break LOCA] analysis licensing basis

for FNP [Farley Nuclear Plant]. All

acceptance criteria (including LOCA peak

clad temperature, DNB criteria, containment

temperature and pressure, and dose limits)

continue to be met. Therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

DNB PARAMETERS CHANGES

* * * *

1. The proposed technical specification

changes for DNB parameters do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated in the FNP FSAR. The mechanical

design features associated with VANTAGE 5

fuel and the improved methodologies (such

as Revised Thermal Design Procedure)

provide capability for relaxation of analytical

input parameters such that increased DNBR

[DNB ratio] margin can be generated without

violation of any acceptance criteria. The

indicated DNB parameters bound the

analytical values used to support the

proposed uprating. In each case, the

appropriate design and acceptance criteria

are met. All performance requirements for

any system or component have been

evaluated and support the revised analysis

assumptions. Overall plant integrity is not

reduced. Furthermore, the parameter changes

are associated with features used as limits or

mitigators to assumed accident scenarios and

are not accident initiators. Therefore, the

probability of an accident has not

significantly increased.
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52589

The radiological consequences of accidents

previously evaluated in the FSAR have been

assessed due to the proposed technical

specification changes. Evaluations have

confirmed that the doses remain within

previously approved acceptable limits as

well as those defined by 10 CFR [Part] 100.

Therefore, the radiological consequences to

the public resulting from any accident

previously evaluated in the FSAR has not

significantly increased.

2. The proposed technical specification

changes do not create the possibility of a new

or different kind of accident from any

previously evaluated in the FSAR. No new

accident scenarios, failure mechanisms, or

limiting single failures are introduced as a

result of the revised DNB parameters. The

revised analytical assumptions have no

adverse effect and do not challenge the

performance of any other safety-related

system. This has been verified in WCAP

12771, Rev. 1. Therefore, the possibility of a

new or different kind of accident is not

created.

3. The proposed technical specification

changes do not involve a significant

reduction in the margin of safety. The margin

of safety for fuel-related parameters (such as

DNB and Kw/ft) are defined in the Bases to

the Technical Specifications. The

uncertainties associated with the proposed

DNB parameter changes are included in the

core safety limits. Performance of analyses

and evaluations with the reactor core safety

limits defined by RTDP [Revised Thermal

Design Procedure] have confirmed that the

operating envelope defined by the Technical

Specifications continues to be bounded by

the revised analytical basis, which in no case

exceeds the acceptance limits. Therefore, the

margin of safety provided by the analyses in

accordance with these acceptance limits is

not reduced.

MISCELLANEOUS OPERATION AND

MARGIN ENHANCEMENT CHANGES

* * * *

1. The proposed changes do not increase

the probability or consequences of an

accident previously evaluated in the FSAR.

Explicit modeling of these parameters is

included in the uprate analyses and

evaluations. The comprehensive analytical

effort performed to support the proposed

uprating has included a review and

evaluation of all components and systems

(including interface systems and control

systems) that could be affected by this

change. In addition LOCA and non-LOCA

analyses and evaluations have verified that

all acceptance criteria continue to be met. All

systems will function as designed. None of

these proposed changes can directly initiate

any accidents; therefore, the probability of an

accident has not been increased. All dose

consequences have been analyzed or

evaluated with respect to these parameters,

and all acceptance criteria continue to be

met. Therefore, the consequences of an

accident previously evaluated in the FSAR

have not increased.

2. The proposed changes do not create the

possibility of a new or different kind of

accident than any accident already evaluated

in the FSAR. No new accident scenarios,

failure mechanisms or limiting single failures

are introduced as a result of the proposed

changes. The proposed technical

specification changes have no adverse effects

on any safety-related system and do not

challenge the performance or integrity of any

safety-related system. Therefore, the

possibility of a new or different kind of

accident is not created.

3. The proposed technical specification

changes do not involve a significant

reduction in a margin of safety. All analyses

supporting the proposed power uprate reflect

these proposed values. All acceptance

criteria (including LOCA peak clad

temperature, DNB criteria, containment

temperature and pressure, and dose limits)

continue to be met. Therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

ALLOWABLE VALUES AND TRIP

SETPOINTS FOR REACTOR TRIP SYSTEM

AND ESFAS [ENGINEERED SAFETY

FEATURE ACTUATION SYSTEM]

* * * *

1. The proposed changes do not increase

the probability or consequences of an

accident previously evaluated in the FSAR.

The comprehensive engineering effort

performed to support the proposed uprating

has included evaluations or reanalysis of all

accident analyses including all dose related

events. Setpoint calculations have verified

acceptability of the proposed setpoints and

allowable value changes. All systems will

function as designed, and all performance

requirements on these systems have been

verified to be acceptable. Neither allowable

values nor the setpoints initiate any accident;

therefore, the probability of an accident has

not been increased. All dose consequences

have been analyzed or evaluated with respect

to these parameters, and all acceptance

criteria continue to be met. Therefore the

consequences of an accident previously

evaluated in the FSAR have not increased.

2. The proposed setpoints and allowable

value changes do not create the possibility of

a new or different kind of accident than any

accident already evaluated in the FSAR. No

new accident scenarios, failure mechanisms

or limiting single failures are introduced as

a result of the proposed changes. The

proposed technical specification changes

have no adverse effects on any safety-related

system and do not challenge the performance

of integrity of any safety-related system. The

specified trip setpoints associated with the

respective RTS [Reactor Trip System] and

ESFAS functions ensure all accident analyses

criteria continue to be met. Therefore, the

possibility of a new or different kind of

accident is not created.

3. The proposed technical specification

changes do not involve a significant

reduction in a margin of safety. All analyses

supporting the proposed power uprate reflect

these proposed values. Setpoint calculations

demonstrate that margin exists between the

setpoint and the corresponding safety

analysis limits. The calculations are based on

FNP instrumentation and calibration/

functional test methods and include

allowances for uprated power conditions. All

acceptance criteria (including LOCA peak

clad temperature, DNB criteria, containment

temperature and pressure, and dose limits)

continue to be met. 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request:

September 17, 1997 (TS 97-02)

Description of amendment request:

The proposed changes would revise

Section 4.6.2.1 of the Sequoyah

Technical Specifications (TS) to change

the parameters to be monitored during

the inservice inspection surveillance

testing of the containment spray system

pumps. The changes would also adopt

provisions in the Westinghouse

Improved Standard TS (NUREG-1431)

that affect that section of the Sequoyah

TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revisions to the containment

spray system surveillances for the pumps,

valves, and nozzles do not change the intent

of the current TS requirements. These

revisions only affect the TS operability

testing requirements without changing the

system functions. These functions are not

considered to be accident initiators. The

proposed surveillance wording is not based

on changes to the plant although a

modification to flow orifices for the

containment spray pumps created the need to

revise the surveillance that verifies pump

developed head. The revisions primarily

provide flexibility for required methods to

verify system operability as well as utilizing

less prescriptive operability limits and

conditions for testing. The testing flexibility

and less prescriptive requirements do not
52590 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

relax the intent to properly verify operability

of the containment spray system but do allow

for changes in testing that continue to ensure

the appropriate operability requirements.

Since these revisions are not directly related

to modifications of the plant or result in

different methods for operating the plant,

there is no change that could increase the

probability of an accident. In addition, the

consequences of an accident are not

increased because there has not been a

change that would impact the safety

functions of the containment spray system.

These revisions will continue to properly

verify the operability of the containment

spray system.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The containment spray system functions

are not changed as discussed above and the

operating practices for the plant remain the

same. The testing methods can be modified

as a result of the proposed revisions but will

continue to maintain appropriate

verifications of system operability. These

testing methods as well as the containment

spray system are not considered to be a

potential initiator of accidents. Therefore,

these revisions will not impact the operation

of systems that could initiate an accident and

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 revisions do not directly

change the limits for containment spray

system operability although they do provide

the flexibility to properly revise limits

resulting from system modifications. This

type of limit revision would be necessary to

adequately verify system operability. The

appropriate limits continue to be required by

the proposed TS surveillance requirements.

Therefore, the proposed revisions do not

allow inappropriate changes to setpoints or

operating requirements that maintain the

margin of safety and no reduction in this

margin is involved in this request.

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

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: August

26, 1997

Description of amendment request:

The proposed amendment would

change Technical Specification (TS) 3/

4.2, ‘‘Power Distribution Limits.’’ The

DNB Parameters Limiting Condition for

Operation would be modified consistent

with an industry notification.

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 Davis-Besse Nuclear Power Station has

reviewed the proposed changes and

determined that a significant hazards

consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no accident initiators,

assumptions or probabilities are affected by

the proposed change. The proposed change

corrects a nonconservative Technical

Specification Action statement by removing

provisions which allow continued Mode 1

plant operation in the event the Reactor

Coolant System flow rate is less than the

required value. Under the proposed change,

a power reduction to less than 5 percent of

rated thermal power (Mode 2) will be

required if the Reactor Coolant System flow

rate is less than the required Technical

Specification value.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed change does

not affect any equipment, accident

conditions, or assumptions which could lead

to a significant increase in radiological

consequences of an accident. The proposed

change will ensure accident analyses remain

valid if the Reactor Coolant System flow rate

becomes less than the required value.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new

accident initiators will be introduced by the

proposed change. No equipment or

operations will be affected.

3. Not involve a significant reduction in a

margin of safety because under the proposed

Technical Specification Action statement a

power reduction to less than 5 percent of

rated thermal power (Mode 2) will be

required if degraded Reactor Coolant System

flow develops. The proposed Action

statement ensures accident analyses’

assumptions are maintained.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: August

22, 1997, as supplemented by letter

dated September 18, 1997

Description of amendment request:

The proposed amendment would revise

the Vermont Yankee Technical

Specifications (TSs) to address the new

low pressure C02 suppression system for

the East and West Switchgear Rooms

and more clearly describe the separation

of the rooms.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated:

The proposed changes support the use of

a newly installed low pressure C02

suppression system for the East and West

Switchgear Rooms, to meet the C02

concentration requirements of NFPA 12

(1993) following detection of a fire condition

in one of the associated rooms. The new low

pressure C02 system consists of a 6 ton

storage tank, piping, valves, associated

instrumentation and controls.

The FSAR [Final Safety Analysis Report]

was reviewed for impact as a result of this

proposed amendment with none being found.

The initiators of the four design basis

accidents, as defined in section 14.6 of the

FSAR, were reviewed with respect to the new

low pressure C02 system. The low pressure

C02 system is not an initiator of any of the

Chapter 14.6 accidents. The low pressure

C0T22 suppression system is classified as a

Non Nuclear Safety (NNS) related system.

However, the C02 dispersion headers have

been seismically mounted to preclude the

possibility of their failure affecting safety

related equipment during a seismic event.

Although the Switchgear Room (East and

West) low pressure C02 system is not used as

a mitigator of any accident listed in section

14.6 of the FSAR, the switchgear contained

in the aforementioned rooms is used to
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52591

mitigate the consequences of the section 14.6

accidents.

The new low pressure C02 system, which

meets NFPA 12 (1993), provides fire

suppression for the affected room by raising

the C02 concentration to a 50% level and

maintains this concentration for a 20 minute

duration upon initiation. As a result, this C02

system prevents a fire in the affected room

from spreading to adjacent rooms and

adversely impacting the adjacent room’s

safety related equipment. Consequently, the

unaffected rooms and associated trains of

equipment remain functional to perform their

intended safety functions if required. The

proposed amendment also reflects the

separation of the switchgear room into two

fire areas with equivalent detection and

suppression.

Based on the above, use of the low pressure

C02 system for East or West Switchgear Room

fire suppression does not create new

initiators, nor degrade the effectiveness of

equipment relied upon to perform mitigative

functions assumed for the previously

evaluated design basis accidents. Therefore,

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any previously

evaluated:

The NNS low pressure C02 system, which

meets NFPA 12 (1993), provides fire

suppression for the East and West Switchgear

Rooms by raising the C02 concentration to a

50% level and maintains this concentration

for a 20 minute duration upon initiation. As

a result, this C02 system prevents a fire in the

affected switchgear rooms from spreading to

adjacent rooms and adversely impacting the

adjacent rooms associated equipment. The

switchgear room is more clearly depicted as

two separate fire areas in the proposed

amendment with equivalent protection. The

C02 suppression header piping located in the

switchgear rooms is seismically supported,

which precludes the possibility of this piping

failing during a seismic event and affecting

safety related equipment located nearby.

The new low pressure C02 system does not

introduce new accident initiators. The low

pressure C02 system is fulfilling the fire

suppression function previously performed

by the existing high pressure C02 system. The

previous separation of the switchgear room

into two separate fire areas, provides

separation of redundant equipment and

equivalent fire detection and suppression for

that equipment. The low pressure C02 system

consists of a 6 ton storage tank, piping,

valves, and associated instrumentation and

controls. There are no failure mechanisms,

associated with the new low pressure C02

equipment, which cannot be categorized

under at least one of the three failure

mechanisms identified in section 14.4.3 of

the FSAR. Consequently, the proposed

amendment will not create the possibility of

a new or different kind of accident from any

previously evaluated.

3. The proposed amendment will not

involve a significant reduction in a margin of

safety.

Technical Specifications 3.13.D/4.13.D

were reviewed with respect to the proposed

amendment to determine if the changes

would result in a reduction in a margin of

safety. The proposed amendment, to allow

use of a low pressure C02 suppression system

for the East or West Switchgear Rooms, does

not degrade the existing fire protection

program. The level of protection provided by

the switchgear room C02 fire protection

system is enhanced by the introduction of the

new low pressure system which meets NFPA

12 (1993) and provides fire suppression for

the East or West Switchgear Rooms by raising

the C02 concentration to a 50% level and

maintains this concentration for a 20 minute

duration upon initiation. Consequently, the

pre-established levels of system operability

in the event of a fire and the assurance of a

safe reactor shutdown, as provided by the fire

protection systems, have not been degraded.

An analysis has been performed to ensure

that either a failure of the low pressure C02

storage tank outside the switchgear rooms, or

a continuous discharge of the entire tank

contents within the switchgear room, will not

adversely affect either control room

habitability or emergency diesel operation.

The designation of separate fire areas for the

switchgear room with equivalent protection

does not decrease safety for this equipment.

As a result, the proposed amendment will

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on

thisreview, 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

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

Ropes and Gray, One International

Place, Boston, MA 02110-2624

NRC Project Director: Ronald B.

Eaton, Acting Project Director

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of amendment request: July 16,

1997

Description of amendment request:

The proposed amendment would add

new minimum reactor vessel pressure

versus reactor vessel metal temperature

(P/T) curves, applicable to 12 EFPY

(effective full power years). These

changes are necessary to support leak

and hydrostatic testing in accordance

with the American Society for

Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code (Code)

Section XI.

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 12 EFPY curve was

developed using the same methodology as

that used in the current 32 EFPY curve and

the 8 EFPY curve. This methodology is

consistent with the guidance provided in

Regulatory Guide 1.99, Revision 2.

Assumptions and parameters were the

same as those used in the 8 EFPY curve

calculation. However, fluence values used in

the calculation were those for 12 EFPY.

Use of the 12 EFPY curves on or before

attainment of 12 EFPY of operation is

equivalent to the previously approved use of

the 32 EFPY curves on or before attainment

of 32 EFPY of operation.

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.

The proposed change introduces no

credible mechanism for unacceptable

radiation release.

The proposed change does not require

physical modification to the plant.

The 12 EFPY curves are consistent with the

previously approved 32 and 8 EFPY curves.

Inservice hydrostatic or leak testing is not

assumed to be an initiator of analyzed events.

Since approval of the proposed amendment

will ensure adequate protection of the reactor

pressure vessel, it will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The accident analyses for the plant as

described in the FSAR are not affected by

this proposed change.

The 12 EFPY curves were developed using

the same methodology as the 32 and 8 EFPY

curves and thus involve no reduction in the

margin of safety as previously evaluated.

The margin of safety, relative to the

available heat sink in the Reactor Coolant

System, is actually increased by the use of

the proposed curves due to the lower allowed

test temperature.

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: Richland Public Library, 955

Northgate Street, Richland, Washington

99352
52592 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

Attorney for licensee: Perry D.

Robinson, Esq., Winston & Strawn, 1400

L Street, N.W., Washington, D.C. 20005-

3502

NRC Project Director: William H.

Bateman

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Carolina Power & Light Company, et

al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of amendments request: August

6, 1997, as supplemented August 26,

1997

Brief description of amendments: The

proposed amendments would address

an unreviewed safety question

associated with handling of the spent

fuel shipping cask at the Brunswick

Steam Electric Plant, Units 1 and 2.Date

of publication of individual notice in

Federal Register: September 17, 1997

(62 FR 48897)

Expiration date of individual notice:

October 17, 1997

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

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 application for amendment:

August 26, 1997

Brief description of amendment

request: The proposed amendments

would approve a modification to the

Diablo Canyon Power Plant, Unit Nos. 1

and 2 auxiliary saltwater (ASW) system

to bypass approximately 800 feet of Unit

1 and 200 feet of Unit 2 Class 1 ASW

pipe, a portion of which is buried below

sea level in the tidal zone outside the

intake structure. This modification was

completed on Unit 1 during the

refueling outage completed this

year.Date of individual notice in

Federal Register: September 16, 1997

(62 FR 48677)

Expiration date of individual notice:

October 16, 1997

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Southern Nuclear Operating Company,

Inc., et al., Docket No. 50-348, Joseph

M. Farley Nuclear Plant, Unit No. 1,

Houston County, Alabama

Date of amendment request:

September 3, 1997

Description of amendment request:

The proposed amendment would allow

a reduction in the number of required

available movable detector thimbles

(flux map paths) for Cycle 15

operation.Date of publication of

individual notice in Federal Register:

September 10, 1997 (62 FR 47695)

Expiration date of individual notice:

October 10, 1997

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

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 amendment request:

September 17, 1997

Description of amendment request:

The proposed amendments would

modify Technical Specification 3/4.4.9,

‘‘Specific Activity,’’ and associated

Bases to reduce the limit associated

with dose equivalent iodine-131.Date of

publication of individual notice in

Federal Register: September 24, 1997

(62 FR 49998)

Expiration date of individual notice:

October 24, 1997

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, OES Nuclear, Inc.,

Pennsylvania Power Company, Toledo

Edison Company, Docket No. 50-440

Perry Nuclear Power Plant, Unit No. 1,

Lake County, Ohio

Date of application for amendment:

August 14, 1997

Brief description of amendment: The

proposed amendment would change the

Perry Nuclear Power Plant design basis

as described in the Updated Safety

Analysis Report. The change will add a

description of the methodology utilized

for determining the systems and

components that are considered to

require protection from tornado

missiles.Date of individual notice in

Federal Register: September 16, 1997

(62 FR 48674).

Expiration date of individual notice:

October 16, 1997

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, OH 44081

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
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52593

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.

Arizona Public Service Company, et al.,

Docket Nos. STN 50-528, STN 50-529,

and STN 50-530, Palo Verde Nuclear

Generating Station, Unit Nos. 1, 2, and

3, Maricopa County, Arizona

Date of application for amendment:

December 27, 1996, as supplemented by

letter dated August 22, 1997

Brief description of amendment: The

amendments change Technical

Specification 3/4.6.1.3.b and its

associated Bases sections to reflect an

increase in the peak containment

internal pressure for the design basis

loss-of-coolant accident (LOCA) from

49.5 psig to 52 psig.

Date of issuance: September 11, 1997

Effective date: September 11, 1997, to

be implemented within 30 days from its

date of issuance.

Amendment No.: Unit 1 - 113; Unit 2

- 106; Unit 3 - 85

Facility Operating License Nos. NPF41, NPF-51, and NPF-74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 21, 1997 (62 FR 27794)

The August 22, 1997, supplemental

letter provided additional clarifying

information and 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 September 11,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Carolina Power & Light Company, et

al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of application for amendments:

July 8, 1997, as supplemented August

22, 1997

Brief description of amendments:

These amendments remove the

suppression chamber water volume

band from Technical Specification

3.6.2.1.a.1 while retaining the

equivalent water level band. The

amendments additionally revised the

volume band to account for the

displacement of water due to the

installation of larger emergency core

cooling system suction strainers.

Date of issuance: September 17, 1997

Effective date: September 17, 1997

Amendment Nos.: 188 and 219

Facility Operating License Nos. DPR71 and DPR-62: Amendments change

the Technical Specifications

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43366)

The August 22, 1997, submittal

provided a correction to the Bases to

reflect a change authorized by a

previous amendment and did not alter

the initial no significant hazards

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

September 17, 1997.No significant

hazards consideration comments

received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

July 1, 1997

Brief description of amendments: The

amendments revise Technical

Specification definition 1.4, Channel

Calibration, to allow an alternative

method of calibrating thermocouples

and resistance temperature detector

sensors. The amendments also make

editorial and administrative corrections

to TS Table 3.3.2-1, Table 3.3.6-1, and

Bases Section 3/4.3.1.

Date of issuance: September 15, 1997

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 102 and 104

Facility Operating License Nos. NPF11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40848)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 15, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Duke Energy Corporation, Docket Nos.

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

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

October 30, 1996, as supplemented by

letters dated April 22, July 2, September

3, and September 4, 1997

Brief description of amendments: The

amendments revise the Reactor Building

Structural Integrity Technical

Specifications regarding the tendon

surveillance program.

Date of Issuance: September 15, 1997

Effective date: The license

amendments are effective as of the date

of issuance and the change to the

facilities shall be implemented prior to

the Unit 1 end-of-cycle 17 outage.

Implementation of the amendments

shall include the provisions that the

licensee provide in the facility Updated

Final Safety Analysis Report

(specifically the Selected Licensee

Commitment Manual) the prescribed

lower limit and the minimum required

value of Reactor Building PostTensioning System tendon forces for

each group of tendons prior to

performing the seventh tendon

surveillance for Unit 1. In addition, the

portion of the Selected Licensee

Commitment Manual related to the

establishment of these limits will be

submitted as soon as available.

Amendment Nos.: 225, 225, 222

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

amendments revised the Technical

Specifications, License Conditions, and

Appendix C.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64383) The April 22, July 2, September

3, and September 4, 1997, letters

provided clarifying information that did

not change the scope of the October 30,

1996, application and the initial

proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 15,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Duke Energy Corporation, Docket Nos.

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

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

June 12, 1997

Brief description of amendments: The

amendments change the name ‘‘Duke

Power Company’’ to ‘‘Duke Energy

Corporation’’ in the Oconee facility

operating licenses and Technical

Specifications.

Date of Issuance: September 16, 1997
52594 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 226, 226, 223

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

amendments revised the Technical

Specifications and Operating Licenses

including Appendix C.

Date of initial notice in Federal

Register: July 2, 1997 (62 FR 35849) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 16, 1997,

and Environmental Assessment dated

August 21, 1997 (62 FR 44495).No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

September 6, 1996, as supplemented

May 23, 1997 and August 13, 1997.

Brief description of amendments:

These amendments revise Item 7.c of

Beaver Valley Power Station, Unit No. 1

(BVPS-1) Technical Specification (TS)

Table 3.3-3 and Item 7.d of Beaver

Valley Power Station, Unit No. 2 (BVPS2) TS Table 3.3-3 to reflect that a safety

injection (SI) signal starts all auxiliary

feedwater (AFW) pumps. The notation

on BVPS-1 TS Table 3.3-5 is revised to

state that the response time is for all

AFW pumps on all SI signal starts.

Items 7.d of BVPS-2 TS Tables 3.3-4 and

4.3-2 is revised to reflect that an SI

signal starts all AFW pumps.

The amendments also revise and

reformat TSs 3/4.7.1.2 to more closely

resemble the wording contained in the

NRC’s ‘‘Standard Technical

Specifications Westinghouse Plant,’’

(NUREG-1431, Revision 1). These

changes require three AFW trains to be

operable and describe what constitutes

an operable train. The mode

applicability for these TSs is expanded

to include Mode 4 when the steam

generator(s) is relied upon for heat

removal.

Date of issuance: September 18, 1997

Effective date: Both units, as of the

date of issuance, to be implemented

within 60 days

Amendment Nos.: 206 and 85

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 19, 1996 (61 FR

58902) The May 23, 1997, and August

13, 1997, letters provided minor

editorial changes that did not change

the initial proposed no significant

hazards consideration determination or

expand the amendment request beyond

the scope of the November 19, 1996,

Federal Register notice. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 18,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of application for amendment:

May 29, 1997

Brief description of amendment: The

amendments consist of changes to the

Technical Specifications (TS) which

correct typographical errors, remove

outdated material, incorporate minor

changes in text, make editorial

corrections, and resolve other

inconsistencies in the Unit 1 and 2 TS.

Date of Issuance: September 22, 1997

Effective Date: September 22, 1997

Amendment Nos.: 152 and 89

Facility Operating License Nos. DPR67 and NPF-16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40849)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 22,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Indian River Community

College Library, 3209 Virginia Avenue,

Fort Pierce, Florida 34981-5596

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: July 8,

1996

Brief description of amendments: The

amendments allowed that the

component cooling water system surge

tank level instrumentation can be

demonstrated operable, by performing a

channel calibration test, during any

plant mode of operation.Date of

issuance: September 23, 1997

Effective date: September 23, 1997, to

be implemented within 30 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 91; Unit 2 -

Amendment No. 78

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 28, 1996 (61 FR 44358)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 23,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

June 19, 1997

Brief description of amendment:

Technical Specification 3/4.7.1.3

requires sufficient water to be available

for the auxiliary feedwater system to

maintain the reactor coolant system at

hot standby for 10 hours before cooling

down to hot shutdown in the next 6

hours. The amendment increases the

required volume of water when the

demineralizer water storage tank and

condensate storage tank are being

credited, makes editorial changes, and

expands the descriptions in Bases

Sections 3/4.7.1.2 and 3/4.7.1.3.

Date of issuance: September 11, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 150

Facility Operating License No. NPF49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40853)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 11, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52595

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

May 7, 1997, as supplemented May 30,

July 29, and September 12, 1997

Brief description of amendments: The

amendments revise Technical

Specification (TS) 3.8, including TS

3.8.D.1 and TS 3.8.D.3, to change TS

limitations on crane operations in the

spent fuel pool enclosure relating to

spent fuel pool special ventilation

system operability. These changes are

necessary to allow movement of loads

over spent fuel stored in the spent fuel

pool enclosure with the spent fuel pool

special ventilation system inoperable.

The staff denied the proposed change to

TS 3.8.D.2. A separate notice of denial

has been sent to the Federal Register for

publication.

Date of issuance: September 15, 1997

Effective date: September 15, 1997,

with full implementation within 30

days. License Condition 4 of Appendix

B is effective immediately upon

issuance of the amendments.

Amendment Nos.: 130 and 122

Facility Operating License Nos. DPR42 and DPR-60: Amendments revised

the Licenses and Technical

Specifications.

Date of initial notice in Federal

Register: July 2, 1997 (62 FR 35850) The

July 29 and September 12, 1997, letters

provided clarifying information within

the scope of the original application and

did not change the staff’s initial

proposed no significant hazards

considerations determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 15,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

July 3, 1997

Brief description of amendment: This

amendment makes changes to Technical

Specification Table 3.6.3-1, ‘‘Primary

Containment Isolation Valves’’ to add

valves to the list, therein.

Date of issuance: September 15, 1997

Effective date: Effective as of the date

of issuance, to be implemented within

60 days.

Amendment No.: 102

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43375)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 15,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

July 7, 1997

Brief description of amendment: The

amendment changes Technical

Specification (TS) 3/4.8.4.2, ‘‘Motor

Operated Valves - Thermal Overload

Protection (BYPASSED),’’ to relocate the

list of applicable valves (TS Table

3.8.4.2-1) to the Hope Creek Generating

Station Updated Final Safety Analysis

Report.

Date of issuance: September 16, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 103

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications and the

License.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43375)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 16, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

April 1, 1997, as supplemented by letter

dated May 30, 1997

Brief description of amendment: The

amendment changed Technical

Specifications (TSs) 4.6.1.1, ‘‘Primary

Containment Integrity;’’ 3/4.6.1.2,

‘‘Primary Containment Leakage;’’ 3/

4.6.1.3, ‘‘Primary Containment Air

Locks;’’ 4.6.1.5.1, ‘‘Primary Containment

Structural Integrity;’’ and 4.6.1.8.2,

‘‘Drywell and Suppression Chamber

Purge System.’’ This amendment also

changed the Bases for 3/4.6.1.2,

‘‘Primary Containment Leakage;’’ 3/

4.6.1.3, ‘‘Primary Containment Air

Locks;’’ 3.4.6.1.5, ‘‘Primary Containment

Structural Integrity;’’ Section 6,

‘‘Administrative Controls;’’ and License

Condition 2.D of Facility Operating

License NPF-57. A new TS, 6.8.4.f,

‘‘Primary Containment Leakage Rate

Testing Program,’’ was added. These

changes modify the TSs and the Facility

Operating License to adopt the

performance based containment leak

rate testing requirements (Option B) of

10 CFR Part 50, Appendix J.Date of

issuance: September 18, 1997

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 104

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications and the

License.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43375)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 18,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

South Carolina Electric & Gas

Company, South Carolina Public

Service Authority, Docket No. 50-395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of application for amendment:

March 26, 1997

Brief description of amendment: The

amendment changes the definition of

‘‘Core Alteration.’’

Date of issuance: September 17, 1997

Effective date: September 17, 1997

Amendment No.: 138

Facility Operating License No. NPF12: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 21, 1997 (62 FR 27800)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 17,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

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:

September 26, 1996, as supplemented

on August 12, 1997 (TS 96-04)

Brief description of amendments: The

amendments change the Technical
52596 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

Specifications (TS) by relocating the fire

protection program details to the

Updated Final Safety Analysis Report

and Fire Protection Plan in accordance

with Generic Letters 86-10 and 88-12.

Date of issuance: September 23, 1997

Effective date: September 23, 1997

Amendment Nos.: 228 and 219

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

TS.

Date of initial notice in Federal

Register: July 2, 1997 (62 FR 35843)The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 23,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402

Tennessee Valley Authority, Docket No.

50-390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendment:

April 30, 1997, as supplemented June

18, July 21 (3 letters), August 7 and 21,

1997

Brief description of amendment: The

proposed amendment would change the

design features section of the Technical

Specifications (TS) to provide for

insertion of Lead Test Assemblies

containing Tritium Producing Burnable

Absorber Rods in the Watts Bar Nuclear

Plant reactor core during Cycle 2.

Date of issuance: September 15, 1997

Effective date: September 15, 1997

Amendment No.: 8

Facility Operating License No. NPF90: Amendment revises the TS.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30644)

The TVA letters dated June 18, July 21,

August 7 and 21, 1997 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 an

Environmental Assessment dated

September 8, 1997, and in a Safety

Evaluation dated September 15,

1997.No significant hazards

consideration comments received:

None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, OES Nuclear, Inc.,

Pennsylvania Power Company, Toledo

Edison Company, Docket No. 50-440

Perry Nuclear Power Plant, Unit 1,

Lake County, Ohio

Date of application for amendment:

January 16, 1996, supplemented

December 6, 1996, and August 15, 1997

Brief description of amendment: The

amendment extended the test interval

for the drywell bypass leakage rate test

from 18 months to 10 years. Also, some

surveillances for the drywell air locks

were increased from 18 months to 24

months.

Date of issuance: September 22, 1997

Effective date: September 22, 1997

Amendment No.: 88

Facility Operating License No. NPF58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 2, 1996 (61 FR 3951)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 22,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, OES Nuclear, Inc.,

Pennsylvania Power Company, Toledo

Edison Company, Docket No. 50-440

Perry Nuclear Power Plant, Unit No. 1,

Lake County, Ohio

Date of application for amendment:

May 2, 1997

Brief description of amendment: The

amendment revises an existing

exception to Limiting Condition for

Operation (LCO) 3.0.4 as it applies to

LCO 3.6.1.9 for the main steam isolation

valve (MSIV) leakage control system

(LCS) by making the exception

permanent and clarifying that it only

applies for the inboard MSIV LCS

subsystem.

Date of issuance: September 24, 1997

Effective date: September 24, 1997

Amendment No.: 89

Facility Operating License No. NPF58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33135)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 24,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2b, Benton County,

Washington

Date of application for amendment:

August 14, 1997

Brief description of amendment: The

amendment revises Technical

Specification (TS) 5.5.6 by adding a note

that would extend the surveillance

interval to perform the inservive testing

(IST) full stroke exercise of primary

containment isolation check valve TIPV-6 until the 1998 refueling outage,

scheduled to begin no later than May

15, 1998, or until a plant shutdown of

sufficient duration occurs to allow TIPV-6 testing, whichever occurs first.

Date of Issuance: September 18, 1997

Effective date: September 18, 1997, to

be implemented within 30 days of

issuance.

Amendment No.: 152

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.Public

comments requested as to proposed no

significant hazards consideration: Yes

(62 FR 45280 dated August 26, 1997).

The notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by September 25,

1997, but indicated that if the

Commission makes a final no significant

hazards consideration determination

any such hearing would take place after

issuance of the amendment. The

Commission’s related evaluation and

final no significant hazards

consideration determination are

contained in a Safety Evaluation dated

September 18, 1997.

Attorney for licensee: Perry D.

Robinson, Esq., Winston & Strawn, 1400

L Street, N.W., Washington, D.C. 20005-

3502

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Wisconsin Electric Power Company,

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

Beach Nuclear Plant, Unit Nos. 1 and

2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

January 16, 1997 (TSCR-191), as

supplemented on April 17, August 7,

and August 27, 1997
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52597

Brief description of amendments:

These amendments increase the

minimum volume and boron

concentration for the refueling water

storage tanks and the boric acid storage

tanks. Additionally, these amendments

increase the minimum concentration of

boric acid in the safety injection

accumulator, the reactor coolant system

during refueling operations, and the

reactor coolant system during positive

reactivity changes made when

containment integrity is not maintained.

Date of issuance: September 23, 1997

Effective date: September 23, 1997,

with full implementation within 45

days

Amendment Nos.: 180 and 184

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

the Technical Specifications.

Date of initial notice in Federal

Register: April 23, 1997 (62 FR 19836)

The April 17, August 7, and August 27,

1997, submittals provided clarifying

information within the scope of the

original application and did not change

the staff’s initial proposed no significant

hazards considerations determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 23,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: The Lester Public Library,

1001 Adams Street, Two Rivers,

Wisconsin 54241

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: July 3,

1997

Brief description of amendment: The

amendment changes the definition for

an alteration of the reactor core to one

that is consistent with the intent of the

Improved Standard Technical

Specifications.

Date of issuance: September 18, 1997

Effective date: September 18, 1997

Amendment No.: 109

Facility Operating License No. NPF42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40861)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 18,

1997.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

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: July 3,

1997

Brief description of amendment: The

amendment modifies Technical

Specifications 5.3.1, ‘‘Fuel Assemblies’’

and 6.1.9.6, ‘‘CORE OPERATING

LIMITS REPORT (COLR)’’ to add ZIRLO

as fuel material and the use of limited

zirconium alloy filler rods in place of

fuel rods.

Date of issuance: September 22, 1997

Effective date: September 22, 1997, to

be implemented within 30 days of

issuance.

Amendment No.: 110

Facility Operating License No. NPF42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40860)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 22,

1997.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 LicensesAnd Final

Determination Of No Significant

Hazards ConsiderationAnd

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

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
52598 Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices

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

November 7, 1997, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

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).

Pennsylvania Power and Light

Company, Docket No. 50-387,

Susquehanna Steam Electric Station,

Unit 1, Luzerne County, Pennsylvania

Date of application for amendment:

September 15, 1997, as supplemented

by letter dated September 16, 1997

Brief description of amendment: The

amendment revised the applicability

requirement in Technical Specifications

(TSs) Sections 3.4.2, ‘‘Safety/Relief

Valves’’ (Action c), 4.4.2, and 3.3.7.5,

‘‘Accident Monitoring Instrumentation’’

(TS Table 3.3.7.5-1, Action 80). The

change to the referenced TSs adds the

following applicability footnote:

Compliance with these requirements for

the ‘‘S’’ SRV acoustic monitor is not

required for the period beginning

September 12, 1997, until the next unit

shutdown of sufficient duration to allow

for containment entry, not to exceed the

10th refueling and inspection outage.

Date of issuance: September 23, 1997

Effective date: September 23, 1997

Amendment No.: 169

Facility Operating License No. NPF14: This amendment revised the

Technical Specifications.Public

comments requested as to proposed no

significant hazards consideration: No.

On September 17, 1997, the staff issued

a Notice of Enforcement Discretion,

which was immediately effective and

remained in effect until this amendment

was issued.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, consultation with the

State of Pennsylvania, and final no

significant hazards consideration

determination are contained in a Safety

Evaluation dated September 23, 1997.
Federal Register / Vol. 62, No. 195 / Wednesday, October 8, 1997 / Notices 52599

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

NRC Project Director: John F. Stolz

Dated at Rockville, Maryland, this 1st day

of October 1997.

For the Nuclear Regulatory Commission

John N. Hannon,

Acting Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 97–26502 Filed 10–7–97; 8:45 am]

BILLING CODE 7590–01–F

NUCLEAR REGULATORY

COMMISSION

Draft Regulatory Guide; Issuance,

Availability

The Nuclear Regulatory Commission

has issued for public comment a

proposed revision of a guide in its

Regulatory Guide Series. This series has

been developed to describe and make

available to the public such information

as methods acceptable to the NRC staff

for implementing specific parts of the

NRC’s regulations, techniques used by

the staff in evaluating specific problems

or postulated accidents, and data

needed by the staff in its review of

applications for permits and licenses.

The draft guide is a proposed

Revision 1 to Regulatory Guide 3.54,

and it is temporarily identified as DG–

3010, ‘‘Spent Fuel Heat Generation in an

Independent Spent Fuel Storage

Installation.’’ The guide is in Division 3,

‘‘Fuels and Materials Facilities.’’ This

regulatory guide is being revised to

present a method that is acceptable to

the NRC staff for calculating heat

generation rates for use as design input

for an independent spent fuel storage

installation. The procedures proposed

in this guide, for both boiling water

reactors and pressurized water reactors,

are simpler and therefore are expected

to be more useful to applicants and

reviewers.

The draft guide has not received

complete staff review and does not

represent an official NRC staff position.

Public comments are being solicited

on the guide. Comments should be

accompanied by supporting data.

Written comments may be submitted to

the Rules and Directives Branch,

Division of Administrative Services,

Office of Administration, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001. Copies of comments

received may be examined at the NRC

Public Document Room, 2120 L Street

NW., Washington, DC. Comments will

be most helpful if received by January

2, 1998.

Although a time limit is given for

comments on this draft guide,

comments and suggestions in

connection with items for inclusion in

guides currently being developed or

improvements in all published guides

are encouraged at any time.

You may also provide comments via

the NRC’s interactive rulemaking

website through the NRC home page

(http://www.nrc.gov). This site provides

the availability to upload comments as

files (any format), if your web browser

supports that function. For information

about the interactive rulemaking

website, contact Ms. Carol Gallagher,

(301) 415–5905; e-mail CAG@nrc.gov.

The draft guide may also be viewed and

downloaded at this interactive site.

Regulatory guides are available for

inspection at the Commission’s Public

Document Room, 2120 L Street NW.,

Washington, DC. Requests for single

copies of draft or final guides (which

may be reproduced) or for placement on

an automatic distribution list for single

copies of future draft guides in specific

divisions should be made in writing to

the U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

Attention: Printing, Graphics and

Distribution Branch; or by fax at (301)

415–5272. Telephone requests cannot be

accommodated. Regulatory guides are

not copyrighted, and Commission

approval is not required to reproduce

them.

(5 U.S.C. 552(a))

Dated at Rockville, Maryland, this 12th day

of September 1997.

For the Nuclear Regulatory Commission.

Joseph A. Murphy,

Director, Division of Regulatory Applications,

Office of Nuclear Regulatory Research.

[FR Doc. 97–26639 Filed 10–7–97; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

to Withdraw From Listing and

Registration; (Anicom, Inc., Common

Stock, $.001 Par Value) File No. 1–

13642

October 1, 1997.

Anicom, Inc. (‘‘Company’’) has filed

an application with the Securities and

Exchange Commission (‘‘Commission’’),

pursuant to Section 12(d) of the

Securities Exchange Act of 1934 (‘‘Act’’)

and Rule 12d2–2(d) promulgated

thereunder, to withdraw the above

specified security (‘‘Security’’) from

listing and registration on the Chicago

Stock Exchange, Inc. (‘‘CHX’’ or

‘‘Exchange’’).

The reasons cited in the application

for withdrawing the Security from

listing and registration include the

following:

The Board of Directors of the

Company believes that maintenance of

the dual listing on both the Chicago

Stock Exchange and the Nasdaq

National Market is not in the best

interests of the Company’s stockholders.

No trading of the Company’s Security

has taken place on the CHX since May

1995. All of the trading activity in the

Security has taken place on the Nasdaq

National Market. Accordingly, the Board

of Directors believes that the costs of

maintaining a listing on the CHX do not

justify the Company’s continued listing

on the CHX given the lack of trading on

the Exchange.

By letter dated September 10, 1997,

the CHX states that the Company has

complied with the Exchange’s rules

relating to the delisting of the

Company’s Security.

Any interested person may, on or

before October 23, 1997, submit by letter

to the Secretary of the Securities and

Exchange Commission, 450 Fifth Street,

NW., Washington, DC 20549, facts

bearing upon whether the application

has been made in accordance with the

rules of the exchange and what terms,

if any, should be imposed by the

Commission for the protection of

investors. The Commission, based on

the information submitted to it, will

issue an order granting the application

after the date mentioned above, unless

the Commission determines to order a

hearing on the matter.

For the Commission, by the Division of

Market Regulation, pursuant to delegated

authority.

Jonathan G. Katz,

Secretary.

[FR Doc. 97–26572 Filed 10–7–97; 8:45 am]

BILLING CODE 8010–01–M

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

to Withdraw From Listing and

Registration; (Apollo Eye Group,

Common Stock, $0.001 Par Value and

Redeemable Warrants) File No. 1–

12812

October 1, 1997.

Apollo Eye Group (‘‘Company’’) has

filed an application with Securities and

Exchange Commission (‘‘Commission’’),