34884 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

to discuss matters the release of which

would constitute a clearly unwarranted

invasion of personal privacy per 5

U.S.C. 552b(c)(6).

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. Sam

Duraiswamy, Chief, Nuclear Reactors

Branch (telephone 301/415–7364),

between 7:30 A.M. and 4:15 P.M. EDT.

ACRS 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: June 27, 1996.

Andrew L. Bates,

Advisory Committee Management Officer.

[FR Doc. 96–16962 Filed 7–2–96; 8:45 am]

BILLING CODE 7590–01–M

UNITED STATES NUCLEAR

REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from June 8, 1996,

through June 21, 1996. The last

biweekly notice was published on June

19, 1996 (61 FR 31171).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By August 2, 1996, 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.
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34885

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: May 1,

1996

Description of amendment request:

The proposed amendment would

modify Table 3.1.1, ‘‘Reactor Protection

System (SCRAM) Instrumentation

Requirement,’’ Table 3.2.C.1,

‘‘Instrumentation that Initiates Rod

Blocks,’’ and Technical Specification 3/

4.4, ‘‘Standby Liquid Control.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Note 7 to Table 3.1.1 and Note 6 to Table

3.2.C.1

The changes to Note 7 to Table 3.1.1 and

the addition of Note 6 to Table 3.2.C.1 are

proposed to clarify their requirements, the

appropriate action to take, and their

relationship to plant modes. This revised

scram and rod block applicability is

acceptable because control rods withdrawn

from a core cell containing no fuel assemblies

have a negligible impact on the reactivity of

the core, and, therefore, these features are not

required to be operable (i.e. provide the

capability to scram). Provided all rods

otherwise remain inserted, the RPS [Reactor

Protection System] functions serve no

purpose and are not required. In this

condition, the required shutdown margin

(Specification 3.3.A.1) and the required onerod-out interlock (Specification 3.10.A)

ensure that no event requiring the RPS or

Rod Block will occur.

The Actions of Table 3.1.1 for inoperable

equipment were previously revised in

Amendment ι147 to be consistent with the

improved STS [Standard Technical

Specifications]. Action (A) requires fully

inserting all insertable control rods in core

cells containing one or more fuel assemblies.

Since Specification 3.10.A requires all

control rods to be fully inserted during fuel

movement, the proposed applicable

conditions cannot be entered while moving

fuel. In addition, Specification 3.10.D used

for controlling multiple control rod removal,

requires all control rods in a 3X3 array

centered on the CRDs [Control Rod Drive]

being removed to be fully inserted and

electrically disarmed and all other control

rods fully inserted. The only possible action

is control rod withdrawal, which is

addressed by Action A.

Hence operating Pilgrim in accordance

with the proposed changes will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Section 3/4.4

The proposed change involves

reformatting, renumbering, and rewording of

the existing Technical Specifications and

Bases along with other changes to the

Technical Specifications discussed above.

The reformatting, renumbering, and

rewording along with the other changes

listed involves no technical changes to

existing Technical Specifications, and does

not impact initiators of analyzed events. It

also does not impact the assumed mitigation

of accidents or transient events. Therefore,

the change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change relocates

requirements to other sections of the

Technical Specifications, to plant

procedures, or to the Technical

Specifications BASES. The procedure change

and BASES change processes require any

changes that reflect plant design as described

in the FSAR [Final Safety Analysis Report] be

evaluated in accordance with 10 CFR 50.59.

Since any changes will be evaluated per 10

CFR 50.59, no increase (significant or

insignificant) in the probability or

consequences of an accident previously

evaluated will be allowed. Therefore, this

change will not involve a significant increase

in the probability or consequences of an

accident previously evaluated.
34886 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

The proposed change provides more

stringent requirements than previously

existed in the Technical Specifications. The

more stringent requirements will not result in

operation that will increase the probability of

initiating an analyzed event. If anything the

new requirements may decrease the

probability or consequences of an analyzed

event by incorporating the more restrictive

changes discussed above. The change will

not alter assumptions relative to mitigation of

an accident or transient event. The more

restrictive requirements will not alter the

operation of process variables, structures,

systems, or components as described in the

safety analyses. Therefore, the change will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change deletes the

requirements for Standby Liquid Control

(SLC) System operability during Hot

Shutdown, Cold Shutdown, and Refueling.

The SLC System is not assumed in the

initiation of any previously evaluated events

and therefore the proposed change will not

increase the probability or consequence of a

previously analyzed accident. The SLC

System is not assumed to operate in the

mitigation of any previously analyzed

accidents which are assumed to occur during

Hot Shutdown, Cold Shutdown or Refueling.

This change will not result in operation that

will increase the probability of initiating an

analyzed event. This change will not alter

assumptions relative to mitigation of an

accident or alter the operation of process

variables, structures, systems, or components

as described in the safety analyses. Therefore,

this change will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change adds an action for

both SLC subsystems inoperable that delays

the requirement to initiate plant shutdown

immediately and allows time to recover at

least one subsystem before subjecting the

plant to a potentially unnecessary transient.

Allowing a short period of time to recover

one subsystem is acceptable because of the

large number of independent control rods

available to shut down the reactor and the

diversity of means available to cause control

rod insertion. This change will not alter

assumptions relative to mitigation of an

accident or alter the operation of process

variables, structures, systems, or components

as described in the safety analyses. Therefore,

this change will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change deletes requirements

for demonstrating operability of the

redundant subsystems which eliminates

excessive and unnecessary testing of safety

significant equipment. This is consistent

with guidance 10.1 of Generic Letter 93-05,

‘‘Line-Item Technical Specifications

Improvements to Reduce Surveillance

Requirement for Testing During Power

Operations’’. The change does not affect the

ability of the SLC system to perform on

demand, and by actually lowering the

number of demands to demonstrate

operability, reduces the probability of

equipment failure. Since the change will not

alter assumptions relative to mitigation of an

accident or alter the operation of process

variables, structures, systems, or components

as described in the safety analyses, the

change will not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed change replaces the

requirement to verify B-10 enrichment

concentration by test anytime boron is added

to the solution and each refueling outage

with verifying the enrichment prior to

addition. Since enrichment of the solution in

the tank cannot change by any other means

but chemical addition, ensuring that only

properly enriched material is available for

addition is adequate to maintain enrichment

at the required level. This change will not

alter assumptions relative to mitigation of an

accident or alter the operation of process

variables, structures, systems, or components

as described in the safety analyses. Therefore,

this change will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Note 7 to Table 3.1.1 and Note 6 to Table

3.2.C.1

The changes to Note 7 to Table 3.1.1, and

the addition of Note 6 to Table 3.2.C.1 are

proposed to clarify their requirements, the

appropriate action to take, and their

relationship to plant modes. This revised

scram and rod block applicability is

acceptable because control rods withdrawn

from a core cell containing no fuel assemblies

have a negligible impact on the reactivity of

the core, and, therefore, are not required to

be operable. Provided all rods otherwise

remain inserted, the RPS functions serve no

purpose and are not required. In this

condition, the required shutdown margin

(Specification 3.3.A.1) and the required onerod-out interlock (Specification 3.10.A)

ensure that no event requiring the RPS or

Rod Block will occur.

The Actions of Table 3.1.1 for inoperable

equipment were previously revised in

Amendment ι147 to be consistent with the

improved STS. Action (A) requires fully

inserting all insertable control rods in core

cells containing one or more fuel assemblies.

Since Specification 3.1O.A requires all

control rods to be fully inserted during fuel

movement, the proposed applicable

conditions cannot be entered while moving

fuel. In addition, Specification 3.10.D, used

for controlling multiple control rod removal,

requires all control rods in a 3X3 array

centered on the CRDs being removed to be

fully inserted and electrically disarmed and

all other control rods fully inserted. The only

possible action is control rod withdrawal,

which is addressed by Action A. Hence,

operating Pilgrim in accordance with the

proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Section 3/4.4

The proposed change involves

reformatting, renumbering, and rewording of

the existing Technical Specifications and

Bases along with other changes to the

Technical Specifications discussed above.

The reformatting, renumbering, and

rewording along with the other changes

listed involves no technical changes to

existing Technical Specifications. These

changes are administrative and do not impact

the assumed mitigation of accidents or

transient events. Therefore, these changes do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change relocates

requirements to other Technical

Specification sections, to plant procedures,

or to the Technical Specification BASES.

Relocating requirements will not alter the

plant configuration (no new or different type

of equipment will be installed) or changes in

methods governing normal plant operation.

Relocating requirements will not impose

different requirements and adequate control

of information will be maintained. Relocating

requirements will not alter assumptions

made in the safety analysis and licensing

basis. Therefore, these changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes make some existing

requirements more restrictive and add

additional requirements to the Technical

Specifications but will not alter the plant

configuration (no new or different type of

equipment will be installed) or change

methods governing normal plant operation.

These changes do impose different

requirements, however, they are consistent

with assumptions made in the safety

analyses. Therefore, these changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change relaxes the modes of

applicability for the SLC. Relaxing the

applicability will not involve a physical

alteration of the plant (no new or different

type of equipment will be installed) or

changes in methods governing normal plant

operation. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety.

Note 7 to Table 3.1.1 and Note 6 to Table

3.2.C.1

This revised scram and rod block

applicability is acceptable because control

rods withdrawn from a core cell containing

no fuel assemblies have a negligible impact

on the reactivity of the core, and, therefore,

are not required to be operable (provide a

scram). Provided all rods otherwise remain

inserted, the RPS functions serve no purpose

and are not required. In this condition, the

required shutdown margin (Specification

3.3.A.1) and the required one-rod-out

interlock (Specification 3.10.A) ensure that

no event requiring the RPS or Rod Block will

occur.

The Actions of Table 3.1.1 for inoperable

equipment were previously revised in
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34887

Amendment ι147 to be consistent with the

improved STS. Action (A) requires fully

inserting all insertable control rods in core

cells containing one or more fuel assemblies.

Since Specification 3.10.A requires all

control rods to be fully inserted during fuel

movement, the proposed applicable

conditions cannot be entered while moving

fuel. In addition, Specification 3.10.D, used

for controlling multiple control rod removal,

requires all control rods in a 3X3 array

centered on the CRDs being removed to be

fully inserted and electrically disarmed and

all other control rods fully inserted. The only

possible action is control rod withdrawal,

which is adequately addressed by Action A.

Therefore, operating Pilgrim in accordance

with the proposed changes will not involve

a significant reduction in a margin of safety.

Section 3/4.4

The administrative changes involve no

technical changes. These proposed changes

will not reduce a margin of safety because

there is no impact on any safety analysis

assumptions. Also, because the change is

administrative in nature, no question of

safety is involved. Therefore, these changes

do not involve a significant reduction in a

margin of safety. The change relocates

requirements to other Technical

Specification sections, to plant procedures,

or to the Technical Specification BASES.

These changes will not reduce a margin of

safety since there is no impact on any safety

analysis assumptions. In addition, the

requirements to be transposed are the same

as the existing Technical Specifications.

Since any changes to plant procedures and

Technical Specification BASES are required

to be evaluated per 10 CFR 50.59, no

reduction (significant or insignificant) in a

margin of safety will be allowed. Therefore,

these changes will not involve a significant

reduction in a margin of safety.

The addition of new requirements and

making existing ones more restrictive either

increases or does not affect the margin of

safety. These changes do not impact any

safety analysis assumptions. As such, no

question of safety is involved. Therefore,

these changes will not involve a significant

reduction in a margin of safety.

The proposed change would remove a

backup (in the Hot Shutdown, Cold

Shutdown, and Refueling Modes) to the

available systems for reactivity control;

however, this backup is not considered in the

margin of safety when determining the

required reactivity for shutdown and

refueling events. This change will have no

impact on any safety analysis assumptions.

As such, no question of safety is involved.

Therefore, this change does not involve a

significant reduction in a margin of safety.

The SLC system is not assumed to function

in any DBA or transient and is not the

primary success path of a safety sequence

analysis. It is a backup to the CRD scram

function, therefore, allowing a short period of

time to recover one subsystem will have no

impact on any safety analysis assumptions.

As such, no question of safety is involved.

Therefore, this change does not involve a

significant reduction in a margin of safety.

The change does not alter the requirements

for enrichment/ concentration of the boron

solution necessary to satisfy 10 CFR 50.62.

Since enrichment of the solution in the tank

cannot change by any other means but

chemical addition, this change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

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:

November 15, 1995

Description of amendments request:

The proposed amendments would

revise the Technical Specifications (TS)

to alter the wording of TS 4.8.2.5.a in

accordance with the guidance of

Generic Letter (GL) 91-09, ‘‘Modification

of Surveillance Interval For The

Electrical Protection Assemblies In

Power Supplies For The Reactor

Protection System.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendments do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated because the proposed

change does not alter the design, function, or

operation of the EPAs [Electrical Protective

Assemblies]. The proposed amendments

modify the surveillance requirement for an

electrical protective device on the Reactor

Protection System [RPS]. The RPS-EPA units

are designed to protect RPS equipment from

abnormal operating voltage or frequency. The

proposed change will preclude the need to

test the RPS-EPA units during power

operation. This will eliminate the potential

for reactor scrams and Group isolations

during performance of the surveillance, thus,

preventing unwarranted challenges to safety

systems. The proposed change does not affect

any accident precursor or initiator. Therefore,

the probability of an accident is not affected

by the proposed change. The proposed

amendments do not affect the operability of

the RPS-EPA units. The proposed change

does not affect the ability of the Reactor

Protection System to maintain the integrity of

the fuel cladding, protect the reactor coolant

pressure boundary, or limit the amount of

energy released to primary containment.

Therefore, the consequences of an accident is

not affected by the proposed change.

2. The proposed amendments do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. As stated above, these proposed

amendments do not alter the design,

functions, or operation of the EPAs. The RPS

relay trip logic remains protected from power

supplies operating with abnormal voltage or

frequency. Additionally, the redundancy of

this protection is not changed.

Thus, the proposed amendments do not

create the possibility of a new or different

kind of accident.

3. The proposed amendments do not

involve a significant reduction in a margin of

safety because the benefit to safety by

reducing the frequency of testing during

power operation and attendant possible

challenges to safety systems more than offsets

any risk to safety from relaxing the

surveillance requirement to test the EPAs

during power operation. The testing of each

EPA channel involves a dead-bus transfer

and the momentary interruption of power

results in a half scram and half isolation.

Generic Letter 91-09 notes that many plants

have encountered problems with the reset of

the half trip resulting in inadvertent scrams

and group isolations that challenge safety

systems during power operation. Eliminating

EPA testing at power operation increases the

margin of safety by eliminating the potential

for trips due to testing that challenge safety

systems. An insignificant reduction in the

margin of safety is introduced by increasing

the test interval up to a maximum of a refuel

cycle which will produce a small increase in

risk that an inoperable EPA would not be

detected. The elimination of potential

challenges to safety systems provides a safety

benefit that offsets the increased risks of

component failure.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro
34888 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

Carolina Power & Light Company,

Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of amendment request: June 6,

1996

Description of amendment request:

The proposed change would revise

technical specifications (TS) Section

4.2.3 to allow the licensee to defer the

ultrasonic inspection of the reactor

coolant pump flywheel for one

operating cycle.

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 safety function of the Reactor Coolant

Pump (RCP) flywheel is to provide a

coastdown period during which the RCPs

would continue to provide reactor coolant

flow to the core after a loss of power to the

RCPs. The maximum loading on the RCP

motor flywheel results from overspeed

following a large break Loss of Coolant

Accident (LOCA). The estimated maximum

obtainable speed in the event of a Reactor

Coolant System (RCS) piping break was

established conservatively, and the proposed

one-time change does not affect that analysis.

The RCP flywheels have been carefully

designed and manufactured from high

quality steel. Twenty-two inspections have

been performed at HBRSEP, Unit No. 2 over

the past 25 years and no indications have

been discovered that would affect the

integrity of the flywheel. The Westinghouse

Owners Group (WOG) has performed an

extensive study documented in WCAP14535, ‘‘Topical Report on Reactor Coolant

Pump Flywheel Inspection Elimination,’’ that

includes an evaluation of industry

experience, a stress and fracture evaluation,

and a risk assessment, and has concluded

that RCP flywheel inspections may be safely

eliminated.

Reduced coastdown times due to a single

failed flywheel would not place the plant in

an unanalyzed condition since a locked rotor

(i.e., an instantaneous coastdown) is

analyzed in the Updated Final Safety

Analysis Report (UFSAR). The proposed

change also does not increase the amount of

radioactive material available for release or

modify any systems used for mitigation of

releases during an accident. Therefore, the

proposed change does not involve an

increase in the probability of 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 will not change the

design, configuration, or method of operation

of the plant. Therefore, the proposed change

will not create the possibility of a new kind

of accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in the margin of

safety.

The RCP flywheels have been carefully

designed and manufactured from high

quality steel. Twenty-two inspections have

been performed at HBRSEP, Unit No. 2 over

the past 25 years and no indications have

been discovered that would affect the

integrity of the flywheel. The Westinghouse

Owners Group (WOG) has performed an

extensive study documented in WCAP14535, ‘‘Topical Report on Reactor Coolant

Pump Flywheel Inspection Elimination,’’ that

includes an evaluation of industry

experience, a stress and fracture evaluation,

and a risk assessment, and has concluded

that RCP flywheel inspections may be safety

eliminated. The proposed change would only

result in a one-time deferral of the scheduled

inspection for one operating cycle. In

consideration of the historical integrity of the

HBRSEP, Unit No. 2 RCP flywheels, the

industry experience, the results of the WOG

study, and the deferral of the risk of RCP

flywheel damage during disassembly and

inspection, we conclude that a one operating

cycle deferral of the scheduled RCP flywheel

inspection will not result in a reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: May 31,

1996

Description of amendment request:

The proposed amendment would

change the plant Technical

Specifications (TS) Table 3.3-7, Seismic

Monitoring Instrumentation, and TS

Table 4.3-4, Seismic Monitoring

Instrumentation Surveillance

Requirements, to correct the location

described for one of the three Triaxial

Peak Accelerograph Recorders.

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 does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

These recorders are passive components

which serve only a recording function. They

can neither initiate an accident nor serve to

mitigate accident consequences. The

proposed change serves only to correct the

location, commensurate with design

documents, for one of the three recorders

described in the Technical Specifications.

Accordingly, this change is administrative in

nature. Therefore, there would be no increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed correction is an

administrative change to correct the location

of a recorder currently described in the

Technical Specifications. No physical

alterations to plant equipment are being

made, and there will be no changes that alter

how any safety-related system performs its

function. Therefore, the proposed changes do

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

Technical Specification Bases 3/4.3.3.3

specify the acceptance level for seismic

instrumentation as ‘‘consistency’’ with the

recommendations of Regulatory Guide 1.12.

Since the regulatory guide states only that

one recorder should be provided at a

‘‘selected location on the reactor piping,’’ it

is not material whether it is installed on Loop

1 versus Loop 2. Therefore, the proposed

change does not affect a margin of safety as

defined in the Bases to the Technical

Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34889

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Dates of amendment request:

December 18, 1995, May 3 and June 11,

1996

Description of amendment request:

The licensee proposed to change the

Turkey Point Units 3 and 4 Technical

Specifications (TS) to uprate the core

thermal output of Turkey Point Units 3

and 4 from 2200 MWt to 2300 MWt. The

proposed TS changes were divided into

eight groups. The submittal included a

‘‘No Significant Hazards’’ evaluation for

each of the eight groups. The groupings

are as follows:

TS changes associated with the

uprated power level, the revised core

safety limits, revised DNB [departure

from nucleate boiling] parameters,

Engineered Safety Features Actuation

System (ESFAS) and reactor trip

setpoint changes, and Reactor Coolant

Pump (RCP) Breaker Position Trip, were

evaluated together. The safety of these

proposed changes were verified by the

accident analyses that were completed

in support of the uprated power.

TS changes associated with reducing

the SI [safety injection] pump discharge

head requirement and increasing usable

volume requirements for the

Demineralized Water Storage Tank

(DWST) and the Condensate Storage

Tank (CST) were addressed together.

TS changes associated with

pressurizer and main steam safety valve

(MSSV) setpoint tolerance increases

were assessed together.

TS changes associated with operation

at reduced power with inoperable

MSSVs were assessed separately.

TS changes associated with the

service period for heatup and cooldown

pressure-temperature limit curves were

assessed together.

The Surveillance Requirement change

for the emergency containment cooling

[ECC] unit operability was handled

separately since this was a design

change that required extensive

evaluations.

TS change associated with the methyl

iodide removal efficiency in the Control

Room Emergency Ventilation System

was assessed separately.

All LOCA [loss-of-coolant accident]

related changes dealing with the

peaking factor increase, COLR [core

operating limit report] changes,

Evaluation Model references, and

relocation of peaking factors from the

TS and subsequent inclusion in the

COLR were included in one ‘‘No

Significant Hazards’’ evaluation. All of

the items are closely related since the

LOCA analysis is performed to ensure

peaking factor acceptability.

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.

LICENSE CONDITION, RATED THERMAL

POWER, CORE SAFETY LIMITS, REACTOR

TRIP SYSTEM INSTRUMENTATION TRIP

SETPOINTS, ESFAS INSTRUMENTATION

TRIP SETPOINTS, DNB PARAMETERS AND

RCP BREAKER POSITION TRIP

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes do not involve an

increase in the probability or consequences

of an accident previously evaluated because

operation with these revised values will not

cause any design or analysis acceptance

criteria to be exceeded. The structural and

functional integrity of all plant systems are

unaffected. The overtemperature Delta T and

overpower Delta T reactor trip functions as

well as ESFAS functions are part of the

accident mitigation response and are not

accident initiators. All proposed changes

have been assessed and no design and

analysis acceptance criteria have been

exceeded. Therefore the probability of

occurrence previously evaluated is not

affected.

The proposed changes do not affect the

integrity of the fission product barriers

utilized for mitigation of dose consequences

as a result of an accident. Dose consequences

were reviewed and reanalyzed (as needed)

and found acceptable. Therefore, the

probability or consequences of an accident

previously evaluated are not significantly

increased.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because their effects do not affect

accident initiation sequences. All new

operating configurations have been evaluated

and no new limiting single failures have been

identified. In addition, no new failure modes

have been identified. Therefore, it is

concluded that no new or different kind of

accident from any accident previously

evaluated has been created as a result of

these revisions.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed changes do not involve a

reduction in a margin of safety because the

margin of safety associated with these

parameters as verified by the results of the

accident analyses, are within acceptable

limits. All transients impacted have been

analyzed and have met the applicable

accident analyses acceptance criteria (e.g.,

DNBR [departure from nucleate boiling ratio],

RCS [reactor coolant system] pressure,

secondary side pressure, etc.). The margin of

safety required for each affected safety

analysis is maintained. The adequacy of the

revised Technical Specifications values has

been confirmed such that there is no

reduction in the margin of safety. Therefore,

the proposed changes do not involve a

significant reduction in the margin of safety.

AVAILABLE VOLUME CHANGE FOR

CONDENSATE STORAGE TANK (CST) AND

DEMINERALIZED WATER STORAGE TANK

(DWST), AND REDUCED SAFETY

INJECTION (SI) PUMP DISCHARGE HEAD

REQUIREMENT.

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The revised tank volumes and SI head

requirements have been evaluated with

respect to system performance and analysis

impacts. All accident analysis acceptance

criteria continue to be met. The design

function of all affected systems have been

reviewed and all system design criteria

continue to be met. The structural and

functional integrity of the affected systems

are unaffected. These changes are not

initiators for any accident and therefore the

probability of occurrence of an accident

previously evaluated has not increased.

The proposed changes do not affect the

integrity of the fission product barriers for

mitigation of dose consequences. All dose

consequences remain well within the 10 CFR

100 limits. Therefore there is no increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The revised tank volumes and SI head

requirements do not create the possibility of

a new or different kind of accident from any

accident previously evaluated because these

modifications do not affect accident

initiation sequences. No new operating

configuration is being imposed by the

adjustments that would create a new failure

scenario. In addition, no new failure modes

or limiting single failures have been

identified. Therefore, it is concluded that no

new or different kind of accident from any

accident previously evaluated have been

created as a result of these revisions.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed changes do not involve a

reduction in a margin of safety because the

margin of safety associated with these

parameters, as verified by the results of the

accident analyses and system evaluations, are

within acceptance limits. The margin of

safety required for each affected safety

analysis is maintained. Therefore, the

proposed changes do not involve a

significant reduction in the margin of safety.

PRESSURIZER AND MAIN STEAM

SAFETY VALVE SETPOINT TOLERANCES
34890 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The revised tolerances for main steam

safety valves and pressurizer safety valves do

not involve an increase in the probability or

consequences of an accident previously

evaluated because operation with these

revised values will not cause any design or

analytical acceptance criteria, such as those

applicable to primary and secondary side

pressures to be exceeded. The structural and

functional integrity of the valves are

unaffected by this proposed change. The

tolerance changes do not initiate or cause

initiation of any transient. Therefore, the

probability of occurrence previously

evaluated is not affected.

The changes do not affect the integrity of

the fission product barriers utilized for dose

consequence mitigation. Therefore, the

probability or consequences of an accident

previously evaluated is not increased.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The revised valve tolerances do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated because the tolerances do not affect

accident initiation sequences. No new

operating configuration is being imposed by

the tolerances that would create a new failure

scenario. In addition, no new failure modes

or limiting single failures have been

identified. Therefore, it is concluded that no

new or different kind of accident from any

accident previously evaluated have been

created as a result of these revisions.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The changes to valve tolerances do not

involve a reduction in a margin of safety

because the margin of safety associated with

the MSSVs and the pressurizer safety valves,

as verified by the results of the accident

analyses and valve evaluations, are within

acceptable limits. Transients impacted by

this change have been analyzed and have met

the applicable accident analyses acceptance

criteria, such as those applicable to primary

and secondary side pressure. The margin of

safety required for each affected safety

analysis is maintained. This conclusion is

not changed by the valve tolerances for the

main steam safety valves and the pressurizer

safety valves. Therefore, the changes do not

involve a significant reduction in the margin

of safety.

OPERATION AT REDUCED POWER WITH

INOPERABLE MAIN STEAM SAFETY

VALVES

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The proposed maximum allowable

power level values will ensure that the

secondary side steam pressure will not

exceed 110 percent of the design pressure

following a Loss of Load/Turbine Trip event,

when one or more main steam safety valves

(MSSVs) are declared inoperable. The

proposed change will not impact the

classification of the Loss of Load/Turbine

Trip event as a Condition II probability event

(faults of moderate frequency) per ANSI -

N18.2, 1973. Accordingly, since the proposed

maximum allowable power level will

maintain the capability of the MSSVs to

perform their pressure relief function

associated with a Loss of Load/Turbine Trip

event, there will be no effect on the

probability or consequences of an accident

previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed changes do not

involve any change to the configuration of

any plant equipment, and no new failure

modes have been defined for any plant

system or component. The proposed

maximum allowable power level as specified

in TS Table 3.7-1 will improve the capability

of the MSSVs to perform their pressure relief

function to ensure the secondary side steam

pressure does not exceed 110 percent of

design pressure following a Loss of Load/

Turbine Trip event. Therefore, since the

function of the MSSVs is improved by the

proposed changes, the possibility of a new or

different kind of accident from any accident

previously evaluated is not created.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed changes to the Technical

Specifications do not involve a significant

reduction in a margin of safety. The

algorithm methodology used to calculate the

maximum allowable power level is

conservative and bounding since it is based

on a number of inoperable MSSVs per loop;

i.e., if only one MSSV in one loop is out of

service, the required action to reduce power

to the maximum allowable power level

would be the same as if one MSSV in each

loop were out of service. Another

conservatism with the algorithm

methodology is with the assumed minimum

total steam flow rate capability of the

operable MSSVs. The assumption is that if

one or more MSSVs are inoperable per loop,

the inoperable MSSVs are the largest capacity

MSSVs, regardless of which capacity MSSVs

are actually inoperable.

Therefore, since the maximum allowable

power level calculated for the proposed

changes using the algorithm methodology are

more conservative and ensure that 110

percent of secondary side steam pressure is

not exceeded following a Loss of Load/

Turbine Trip event, this proposed license

amendment will not involve a significant

reduction in a margin of safety.

SERVICE PERIOD FOR HEATUP AND

COOLDOWN PRESSURE-TEMPERATURE

LIMIT CURVES

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Calculation of the service period for the

heatup and cooldown curves does not

involve an increase in the probability or

consequences of an accident previously

evaluated because the calculations were

completed to verify the adequacy of the

existing curves and to determine an

appropriate service period. The use of

approved methods and the acceptable results

have shown that no design or analysis

criteria are changed. The structural and

functional integrity of the reactor vessel has

been verified.

No fission product barriers or inputs to

dose analyses are adversely affected by these

calculations and reverification of the existing

heatup/cooldown curves. Therefore, the

probability or consequences of an accident

previously evaluated are not increased.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The revised service period does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated because the recalculation of an

acceptable service period does not affect

accident initiation sequences. No new

operating configuration is being imposed by

the calculations that would create a new

failure scenario. In addition, no new failure

modes or limiting single failures have been

identified. Therefore, the types of accidents

defined in the UFSAR continue to represent

the credible spectrum of events to be

analyzed which determine safe plant

operation. Therefore, it is concluded that no

new or different kind of accident from any

accident previously evaluated have been

created as a result of these revisions.

(3) Operation of the facility in accordance

with the proposed license amendments

would not involve a significant reduction in

a margin of safety.

Calculations were performed to determine

the service period appropriate for the existing

curves. The changes to service period do not

involve a reduction in a margin of safety

because the margin of safety associated with

the heatup/cooldown curves, as verified by

the results of the analyses, are unchanged.

Therefore, the proposed change to the service

period does not involve a significant

reduction in the margin of safety.

MODIFICATION TO SURVEILLANCE

REQUIREMENT FOR EMERGENCY

CONTAINMENT COOLING SYSTEM

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The purpose of the ECC units is to help

mitigate the consequences of an accident

(i.e., to help maintain the containment

pressure and temperature within their design
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34891

values following a design basis accident).

The ECC units do not operate during normal

operation of the plant. Failure of the ECC

units would not initiate a plant transient or

accident. Therefore, the proposed change

involving the ECC units would not affect the

probability of occurrence of an accident

previously evaluated.

Evaluations demonstrate that, with two

ECC units operating during a LOCA or MSLB

[main steamline break], the containment

pressure and temperature will be maintained

within their design values. These evaluations

also demonstrate that, with two ECC units

operating during a LOCA or MSLB, the

temperature of the CCWS [component

cooling water system] will be maintained

within its design temperature. Therefore, the

proposed change involving the ECC units

would not affect the consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The purpose of the ECC units is to mitigate

design basis accidents, and failure of the ECC

units would not cause a plant transient or

accident. Furthermore, a single failure of an

ECC unit during a LOCA or MSLB would not

lead to a new or different kind of accident.

Although the revised Technical

Specifications require two ECC units to start

automatically on a LOCA signal, they would

also require that all three ECC units be

operable. On a single failure of an operating

ECC unit, there would be sufficient time to

start the standby ECC unit to accomplish the

design function of the ECC system. Therefore,

the proposed amendment would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed change in the actuation logic

of the ECC units would not cause either the

containment pressure and temperature or the

CCWS temperature to exceed their design

values. While the energy released into

containment and subsequently transferred to

the CCWS will increase as a result of the

thermal uprate, this increase is insignificant

and will not result in either the containment

or CCWS exceeding a design limit. Therefore,

the proposed change would not affect the

margin of safety.

CONTROL ROOM EMERGENCY

VENTILATION SYSTEM

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change does not affect the

integrity of the fission product barriers

utilized for mitigation of dose consequences

as a result of an accident. Only the iodide

removal efficiency of the control room

emergency ventilation system is increased,

and this change is in the conservative

direction.

To assure consistency between testing

efficiency and analysis assumptions for postaccident control room doses, the methyl

iodide removal efficiency required to be

demonstrated by laboratory test, is being

increased from 90% to 99%. This increase in

testing efficiency is consistent with the

recommendations set by the NRC staff in

Regulatory Guide 1.52 to support analysis

efficiencies for elemental iodine and methyl

iodide removal of 95%, respectively. Testing

performed to verify methyl iodide removal

efficiency will be performed under

conditions representative of the control room

environment.

Since this change in removal efficiency is

in the conservative direction, plant safety

will not be adversely impacted.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change to the control room

emergency ventilation system iodide removal

efficiency does not create the possibility of a

new or different kind of accident from any

accident previously evaluated because

operation of the control room emergency

ventilation system is not identified in any

accident initiation sequence. The system is

provided to minimize operator exposure to

airborne radioactivity released as a result of

an accident. The new operating configuration

has been evaluated and no new limiting

single failures have been identified as a result

of the proposed modification. Therefore, it is

concluded that no new or different kind of

accidents from any accident previously

evaluated have been created as a result of

these revisions.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed changes do not involve a

reduction in the margin of safety because the

margin of safety associated with this change

is in the conservative direction. Thus, plant

safety will not be adversely impacted and the

margin of safety required for the affected

safety analysis is maintained. The adequacy

of the revised Technical Specification values

to maintain the plant in a safe operating

condition has been confirmed, since the

testing will be done to a more conservative

criteria (i.e., 99% efficiency). Therefore, the

proposed changes do not involve a

significant reduction in the margin of safety.

RELOCATION OF FQ(Z) [HEAT FLUX

HOT CHANNEL FACTOR] AND F Delta H

[NUCLEAR ENTHALPY RISE HOT

CHANNEL FACTOR] LIMITS FROM

TECHNICAL SPECIFICATIONS TO CORE

OPERATING LIMITS REPORT AND

EDITORIAL CORRECTIONS

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The relocation of the values for FQ and F

Delta H from the Technical Specifications to

the Core Operating Limits Report is

administrative in nature and has no impact

on the probability or consequences of any

Design Bases Event (DBE) occurrence which

was previously evaluated. The determination

of the FQ and F Delta H limits will be

performed using methodology approved by

the NRC and poses no significant increase in

the probability or consequences of any

accident previously evaluated.

The changes being proposed as editorial in

nature do not affect assumptions contained

in the safety analyses, the physical design

and/or operation of the plant, nor do they

affect Technical Specifications that preserve

safety analysis assumptions. Therefore, these

proposed changes do not affect the

probability or consequences of accidents

previously analyzed.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The relocation of the FQ and F Delta H

limits from the Technical Specifications to

the Core Operating Limits Report is

administrative in nature and has no impact,

nor does it contribute in any way to the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The determination of the FQ and F Delta H

limits will be performed using NRC-approved

methodology and are submitted to the NRC

as a revision to the COLR to allow the NRC

staff to trend peaking factors. The Technical

Specifications will continue to require

operation within the required core operating

limits and appropriate actions will be taken

if the FQ and F Delta H limits are exceeded.

Therefore, the proposed amendments does

not in any way create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The editorial changes proposed are

administrative in nature and do not affect

assumptions contained in plant safety

analyses, the physical design and/or

operation of the facility, nor do they affect

Technical Specifications that preserve safety

analysis assumptions. Therefore, these

changes do not create the possibility of a new

or different kind of accident.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The relocation of the FQ and F Delta H

limits from the Technical Specifications to

the Core Operating Limits Report is

administrative in nature and has no impact

on the margin of safety. The determination of

the FQ and F Delta H limits will be performed

using methodology approved by the NRC and

does not constitute a significant reduction in

the margin of safety.

The supporting Technical Specification

values are defined by the accident analyses

which are performed to conservatively bound

the operating conditions defined by the

Technical Specifications. Performance of

analysis and evaluation have confirmed that

the operating envelope defined by the

Technical Specifications continues to be

bounded by the analytical basis, which in no

case exceeds the acceptance limits.

Therefore, the margin of safety provided in

the analyses in accordance with the

acceptance limits is maintained and not

significantly reduced.
34892 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

The changes being proposed as editorial in

nature do not relate to or modify the safety

margins defined in, and maintained by the

Technical Specifications. Therefore, the

proposed changes which correct

administrative errors and clarify existing

Technical Specification requirements do not

involve any reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J.

Hebdon

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Dates of amendment request: April

19, 1996, May 10, 1996, and May 28,

1996

Description of amendment request:

The licensee proposed to change the

Turkey Point Units 3 and 4 Technical

Specifications (TS) to address frequency

extension for actions required on a

periodic basis, delete the separate

notification requirement for an

inoperable startup transformer, and

allow the operating RHR loop to be

removed from operation during

refueling operations under certain

conditions.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because the proposed amendments

are purely administrative in nature. These

amendments will not involve a significant

increase in the probability or consequences

of an accident previously evaluated because

they do not affect assumptions contained in

plant safety analyses, the physical design

and/or operation of the plant, nor do they

affect Technical Specifications that preserve

safety analysis assumptions. Therefore, the

proposed changes do not affect the

probability or consequences of accidents

previously analyzed.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The use of the modified specifications can

not create the possibility of a new or different

kind of accident from any previously

evaluated since the proposed amendments

will not change the physical plant or the

modes of plant operation defined in the

facility operating license. No new failure

mode is introduced due to the administrative

changes and clarifications, since the

proposed changes do not involve the

addition or modification of equipment nor do

they alter the design or operation of affected

plant systems, structures, or components.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are unchanged

by the proposed amendments. The modified

specifications which correct administrative

errors and clarify existing Technical

Specification requirements do not

significantly reduce any of the margins of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J.

Hebdon

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: May 30,

1996

Description of amendment request:

The proposed amendment would revise

the technical specifications surveillance

requirement (SR) 3.8.3.4 to specify a 5-

start pressure for the air receivers

associated with the Division III, High

Pressure Core Spray emergency diesel

generator.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequence of

an accident previously evaluated?

The purpose of the proposed Technical

Specification change is to establish

consistency between the basis for the air start

pressure required for the Division I and II

diesels and the value required for the

Division III diesel. The value of 160 psig

currently specified in SR 3.8.3.4 is

representative of a 5-start value for the

Division I and II diesels, however, this value

is not representative of a 5-start for the

Division III diesel. While the 160 psig value

does serve to satisfy the requirements of 10

CFR 50.36 with regard to maintaining the

lowest functional level required for the

Division III diesel to perform its design safety

function, the current value does not serve to

maintain the design margin utilized when

sizing the air receivers for the purpose of

satisfying the Standard Review Plan guidance

contained in section 9.5.6 (NUREG-0800

Revision 2).

The proposed value fully complies with

the guidance provided in NUREG-0800 and

is more conservative than the value currently

included in the Technical Specifications. The

proposed value is well within the capability

of the air system’s design and will not subject

the air system to excessive pressures or

undue cycling of the system’s compressors.

The proposed change has no effect on the

probability of an accident as diesel generators

have no bearing on the initiation of any

analyzed event. In addition, the capability of

the Division III diesel to perform its design

basis function (i.e., starting, accelerating to

rated speed and voltage, and connecting to its

respective bus within 13 seconds) is not

affected by this change. The ability of the

diesel to support the mitigation of analyzed

accidents is not affected and hence the

consequences of any analyzed event are not

affected. Therefore, the proposed change

does not increase the probability or the

consequences of previously analyzed

accidents.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not introduce

any new failure modes. All of the affected

components remain within their applicable

design limits. In addition, the environmental

qualification of any plant equipment is not

adversely affected by the proposed change.

Since the performance of this system is not

adversely affected by this change and the

design margins of this system are not

challenged in a manner differently than

previously analyzed, the proposed change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change raises the required

starting air pressure for the Division III above

that currently required by the Technical

Specifications to establish consistency

between the basis of the Division III value

with the value used for the Division I and II

diesels. Issuance of the proposed change will

establish a 5 start air receiver pressure for

each of the three safety-related diesels at
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34893

River Bend. While the proposed value is

slightly less than the 5 start value discussed

in River Bend’s SER, the proposed value is

supported by the River Bend site-specific test

data and does not adversely affect existing

analyses or system performance. Therefore,

the proposed change does not result in a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC

20005

NRC Project Director: William D.

Beckner

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: June 6,

1996, as supplemented by letters dated

June 7 and 9, 1996

Description of amendment request:

The proposed amendment would revise

the technical specification Limited

Safety System Setting for the MINIMUM

CRITICAL POWER RATIO (MCPR) for

dual recirculation loop operation and

for single recirculation loop operation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

The purpose of the Safety Limit Minimum

Critical Power Ratio (SLMCPR) is to provide

statistical confidence that less than 0.1% of

the fuel rods in a core would experience

transition boiling during the most limiting

analyzed Anticipated Operational

Occurrence (transient). While transition

boiling in a BWR does not in and of itself

signal the onset of fuel cladding failure, this

criterion has been selected as a conservative

and convenient parameter for the evaluation

of fuel designs. Therefore, while this safety

limit does not provide any control over either

the probability or consequences of any

accident previously evaluated, it does ensure

that evaluated transients remain within NRCapproved criteria. Revision of the SLMCPR

will establish in the CNS Technical

Specifications a valid limit, based on the

NRC approved GESTAR II methodology

using cycle-specific inputs. This change will

result in the input of more restrictive core

operating limits into the plant process

computer, ensuring that CNS will be

operated within the constraints of the new

SLMCPR limits of 1.07 for dual recirculation

loop operation, and 1.08 for single

recirculation loop operation. No plant

hardware modifications are associated with

this change. Therefore, since this proposed

change will not change the physical

configuration of the plant, nor result in

operational changes which invalidate

assumptions used in any CNS accident

analysis, this change does not involve an

increase in the probability or consequences

of any accident previously evaluated.

2. Does the proposed License Amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

This change revises the SLMCPR values in

the CNS Technical Specifications in

accordance with a cycle specific analysis

performed for the remainder of the current

cycle. The SLMCPR ensures that less than

0.1% of the fuel rods in a core would

experience transition boiling during the most

limiting Anticipated Operational Occurrence.

Increasing the SLMCPR from 1.06 to 1.07 for

dual recirculation loop operation and from

1.07 to 1.08 for single recirculation loop

operation will ensure that the specified

statistical confidence will be met for all

analyzed transients. This change does not

involve any plant hardware changes. The

only operational changes will be the

institution of appropriate thermal restrictions

on reactor core operation in accordance with

the SLMCPR changes. Therefore, this

proposed change will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Does the proposed change create a

significant reduction in the margin of safety?

This change will establish in the CNS

Technical Specifications, SLMCPR values

that ensure the margin of safety to the NRC

approved Anticipated Operational

Occurrence evaluation acceptance criteria

will be met. Increasing the SLMCPR

institutes more restrictive thermal limitations

on core operation. The change of the

SLMCPR from 1.06 to 1.07 for dual

recirculation loop operation, and from 1.07 to

1.08 for single loop operation will ensure that

the acceptance criteria for evaluated

transients will continue to be met, and that

the appropriate limit is reflected in the CNS

Technical Specifications. Therefore, this

proposed change does not create a reduction

in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Auburn Memorial Library,

1810 Courthouse Avenue, Auburn, NE

68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

NE 68602-0499

NRC Project Director: William D.

Beckner

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of amendment request: May 15,

1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.3.2,

‘‘Isolation Actuation Instrumentation,’’

to establish a range of allowable and trip

setpoints for high temperature (varying

as a function of ambient temperature) in

the Main Steam Line Tunnel Lead

Enclosure Area. Specifically, a new TS

Figure 3.3.2-1 would be added to

provide a curve of allowable

temperature values and a curve of trip

temperature setpoints, both plotted over

a range of ambient temperatures. The

new Figure would be referenced by

Table 3.3.2-2 at item 1.d.3 (High

Temperature Main Steam Line Tunnel

Lead Enclosure Trip Function) by a new

footnote stating:

The trip setpoint and allowable value for

a channel may be established based on Figure

3.3.2-1, if:

a. The actual ambient temperature readings

for all operable channels in the Lead

Enclosure Area are equal to or greater than

the ambient temperature used as the basis for

the setpoint, and

b. The absence of steam leaks in the Main

Steam Line Tunnel Lead Enclosure Area is

verified by visual inspection prior to

increasing a channel setpoint, and

c. A surveillance is implemented in

accordance with Note (d) of Table 4.3.2.1-1.

Similarly, TS Surveillance Table

4.3.2.1-1 would be supplemented at

item 1.d.3 (High Temperature Main

Steam Line Tunnel Lead Enclosure)

with a new footnote stating:

(d) In addition to the normal shift channel

check, if a channel setpoint has been

established using Figure 3.3.2-1, then once

per shift, the actual ambient temperature

reading for all operable channels in the Lead

Enclosure Area shall be verified to be equal

to or greater than the ambient temperature

used as the basis for the setpoint.

Basis for proposed no significant

hazards consideration determination:

The main steam tunnel high

temperature isolation actuation

instrumentation is part of the Leak

Detection System (LDS). It is used to

detect leakage early at 25 gallons per

minute (gpm) and initiate signals to

automatically close the Main Steam

Isolation Valves before a pipe break

could occur. The existing temperature

setpoints for the tunnel lead enclosure

are based upon transient analyses for

steam leaks in the steam tunnel utilizing
34894 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

winter temperatures as an initial

condition. The licensee finds that a

change is needed because actual

temperatures in the tunnel, especially

during the summer, are approaching the

setpoints when steam leakage is not

occurring. Under the present conditions,

a minor disturbance in the turbine

building ventilation system could cause

an unwarranted isolation actuation at

full power with resulting Main Steam

Isolation Valve closure and reactor

scram.

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 operation of NMP2 [Nine Mile Point

Unit 2] in accordance with the proposed

amendment will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The LDS instrumentation in the main

steam line tunnel isolates the Main Steam

Isolation Valves upon sensing a steam leak of

25 gpm. For an elevated ambient temperature

in the Lead Enclosure area, a setpoint

established using the proposed Figure 3.3.2-

1 ensures that the Main Steam Isolation

Valves continue to receive an isolation signal

upon sensing a steam leak of 25 gpm.

Verifying the absence of any steam leak in

the area prior to raising any temperature

instrument setpoint ensures that the ability to

sense a 25 gpm leak is not compromised by

an increased ambient temperature resulting

from a smaller steam leak. The periodic

surveillance to verify the actual ambient

temperature ensures the continued validity of

the ambient temperature used for the setpoint

basis, and provides sufficient advance

indication to take appropriate compensatory

action. Accordingly, this change will not

involve a significant increase in the

consequences of any accident previously

evaluated.

Furthermore, the LDS function provides a

mitigation action for a postulated main steam

line pipe leak which could lead to a pipe

break. This function does not affect any

accident precursors, and the proposed

change does not affect the function of the

LDS system. Accordingly, this change will

not involve a significant increase in the

probability of any accident previously

evaluated.

2. The operation of NMP2 in accordance

with the proposed amendment will not create

the possibility of a new or different kind of

accident from any previously evaluated.

The qualification of safety-related

equipment in the main steam lead enclosure

is evaluated using actual temperatures and

component qualified life is adjusted

accordingly. The temperature elements are

the only safety-related equipment affected by

this change, therefore, the instrumentation

response to previously evaluated accidents

will not be adversely affected. This change

will not affect the performance of safety

related structures. Accordingly, the design

capabilities of those structures, systems and

components affected by the proposed change

are not challenged in a manner not

previously evaluated so as to create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The operation of NMP2 in accordance

with the proposed amendment will not

involve a significant reduction in a margin of

safety.

The proposed change provides a range of

setpoints and allowable values for the Main

Steam Line Tunnel Lead Enclosure

temperatures. The calculation of the

allowable values and trip setpoints was

performed using the same methodologies as

previously employed. For an elevated

ambient temperature in the Lead Enclosure

area, a setpoint established using the

proposed Figure 3.3.2-1 ensures that the

Main Steam Isolation Valves receive an

isolation signal upon sensing a steam leak of

25 gpm, resulting in a main steam line

isolation prior to a pipe break. Therefore, the

proposed change provides the same level of

protection against a main steam line break as

the existing setpoint values. The proposed

setpoints will provide increased scram

avoidance, and thereby reduce unnecessary

challenges to the plant shutdown systems.

Accordingly, the proposed change does 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005-3502

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company,

Dockets Nos. 50-277 and 50-278, Peach

Bottom Atomic Power Station, Units

Nos. 2 and 3, York County,

Pennsylvania

Date of application for amendments:

March 25, 1996

Description of amendment request:

These amendments revise the safety

limit minimum critcal power ratios

(SLMCPRs) to support use of GE-13 fuel

at Peach Bottom Atomic Power Station.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1)The proposed TS [technical

specification] changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The derivation of the revised GE13

SLMCPRs for incorporation into the TS, and

its use to determine cycle-specific thermal

limits, have been performed using USNRC

[U.S. Nuclear Regulatory Commission]-

approved methods within the existing fuel

licensing criteria as discussed in NEDE32198P, ‘‘GE13 Compliance With

Amendment 22 of NEDE-24011-P-A

(GESTAR II),’’ and cannot increase the

probability or severity of an accident.

The basis of the SLMCPRs calculation is to

ensure that greater than 99.9% of all fuel rods

in the core avoid boiling transition if the

limit is not violated. The new SLMCPRs

preserve the existing margin to transition

boiling and fuel damage in the event of a

postulated accident. The fuel licensing

acceptance criteria for the SLMCPRs

calculation apply to the GE13 fuel in the

same manner that they have applied to

previous fuel designs. The probability of fuel

damage is not increased. Therefore, the

proposed TS changes do not involve an

increase in the probability or consequences

of an accident previously evaluated.

2) The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The SLMCPR for the GE13 fuel design is

a Technical Specification numerical value,

designed to ensure that transition boiling

does not occur in 99.9% of all fuel rods in

the core during the limiting postulated

accident. It cannot create the possibility of

any new type of accident. The new SLMCPRs

are calculated using USNRC-approved

methods and have the same calculational

basis as the SLMCPR for other GE fuel

designs previously used at PBAPS, Units 2

and 3. Therefore, the proposed TS changes

do not create the possibility of a new or

different kind of accident, from any accident

previously evaluated.

3) The proposed TS changes do not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the TS

Bases will remain the same. The new

SLMCPRs are calculated using USNRCapproved methods which are in accordance

with the current fuel licensing criteria. The

SLMCPRs for the GE13 fuel remain high

enough to ensure that greater than 99.9% of

all fuel rods in the core will avoid boiling

transition if the limit is not violated, thereby

preserving the fuel cladding integrity.

Therefore, the proposed TS changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34895

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101

NRC Project Director: John F. Stolz

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

No. 50-277, Peach Bottom Atomic

Power Station, Unit No. 2, York County,

Pennsylvania

Date of application for amendment:

June 13, 1996

Description of amendment request:

The proposed amendment to the

Technical Specifications (TS) will

permit a one time performance of

Surveillance Requirement 3.3.1.1.12, for

the Average Power Range Monitor Flow

Biased High Scram function, with a

delayed entry into its associated TS

Conditions and Required Actions for up

to 6 hours provided core flow is

maintained at or above 82 percent. This

change would be in effect until the end

of refueling outage 2R11, currently

scheduled for early October 1996.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

i) The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The APRM system provides monitoring

and accident mitigation functions to limit

peak flux in the core during startup and run

modes. This proposed TS change for delaying

entry into Conditions and Required Actions

associated with SR 3.3.1.1.12 for the APRM

flow bias function will have no impact on the

APRM system or any system that interfaces

with it. No pressure boundary interfaces or

process control parameters will be

challenged.

This change does not affect the operation

of any equipment. Delaying entry into

Conditions and Required Actions associated

with SR 3.3.1.1.12 does not affect either the

initiator of any accident previously evaluated

or any equipment required to mitigate the

consequences of an accident, or the isotopic

inventory in the fuel. Thus, the change does

not increase either the probability or the

consequences of accidents previously

evaluated.

ii) The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Because there is no direct pressure

boundary interface or process control

function associated with the APRM system or

its interfacing electronics, the possibility of a

new or different type of accident than any

previously evaluated will not be created.

Although the flow bias instrument loop does

employ flow transmitters to measure

recirculation drive flow, delaying entry into

Conditions and Required Actions associated

with SR 3.3.1.1.12 will have no impact on

their pressure boundary function. Also,

failure of the sensing line associated with

these transmitters has already been

accounted for in the initial plant design by

including excess flow check valves for

sensing line break isolation.

The proposed change does not introduce a

new mode of plant operation and does not

involve the installation of any new

equipment or modifications to the plant.

Therefore, it does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

iii)The proposed change does not involve

a significant reduction in a margin of safety.

The APRM flow biased high scram

function is not specifically credited in the

safety analysis. However, it is intended to

provide an additional margin of protection

from transient induced fuel damage during

operation where recirculation flow is

reduced to below the minimum required for

rated power operation.

The margin of safety associated with this

change refers to the margin inherent in the

accident analyses that takes credit for the

clamped high flux scram only (i.e., margin

between scramming at 120% peak flux and

the peak flux necessary for fuel damage). The

current reactor operating state (end of cycle

coast down extended core flow) dictates that

only the 120% flux trip be enforced. This trip

remains functional during the APRM flow

biased high scram calibration.

Currently, the Conditions and Required

Actions associated with SR 3.3.1.1.12 permit

a one hour delay prior to entry because it

minimizes risk while allowing time for

restoration or tripping of channels by

operations personnel. Because the APRM

flow biased function is not enforced during

end of cycle, coast down, extended core flow

conditions, extending entry in associated

Conditions and Required Actions from one to

six hours has no impact on the margin

associated with the clamped high flux scram.

In the event core flow drops below 82%, the

flow point below which APRM setpoints

automatically become flow biased, the

associated Conditions and Required Actions

will be entered.

Therefore, extending entry into associated

Conditions and Required Actions associated

with SR 3.3.1.1.12, provided core flow

remains at or above 82%, from one to six

hours does not reduce any margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: May 16,

1996

Description of amendment request:

The proposed amendment to the James

A. FitzPatrick Technical Specifications

(TSs) proposes to delete the requirement

for the Plant Operating Review

Committee (PORC) to review the fire

protection program and implementing

procedures. This proposal will reduce

the administrative burden on the

committee while making PORC’s

responsibilities more consistent with

the other responsibilities described in

Section 6.1.5.6 of the 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:

Operation of the FitzPatrick plant in

accordance with the proposed amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes delete the Plant

Operating Review Committee (PORC) review

of the fire protection program and

implementing procedures, and deleted fire

protection inspection and audit requirements

that are redundant to those performed under

the cognizance of the Safety Review

Committee (SRC). The changes do not

introduce any new modes of plant operation,

make any physical changes, or alter any

operational setpoints. Therefore, the changes

do not degrade the performance of any safety

system assumed to function in the accident

analysis. Consequently, there is no effect on

the probability or consequences of an

accident.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated.

No physical changes to the plant or

changes to equipment operating procedures

are proposed. The changes are administrative

and will not have any direct affect on

equipment important to safety. Therefore the

changes cannot create the possibility of a

new or different kind of accident.
34896 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

3. Involve a significant reduction in the

margin of safety.

Adequacy of the fire protection program

and implementing procedures is assured by

the fire protection license condition, the

procedure review and approval process

implemented by Amendment 222, the

provisions of 10 CFR 50.59, and inspections

and audits performed under the cognizance

of the SRC. Consequently, deleting PORC’s

responsibility for review of the fire protection

program and implementing procedures, and

deleting the inspection and audit

requirements contained in Specification

6.14.A and 6.14.B will not degrade the fire

protection program. Therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: May 30,

1996

Description of amendment request:

The proposed amendment would revise

Minimum Critical Power Ratio Safety

Limit and associated basis. The changes

are required to support introduction of

General Electric Company supplied,

GE12, 10x10 fuel into the Cycle 13

reactor core.

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:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

A change in the SLMCPR [Safety Limit

Minimum Critical Power Ratio] does not

affect initiation of any accident. Operation in

accordance with the revised SLMCPR

ensures the consequences of previously

analyzed accidents are not changed.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The SLMCPR establishes a performance

limit for the fuel. Therefore changing the

limit will not initiate any accident.

3. Involve a significant margin of safety

because:

The analyses performed to determine the

revised SLMCPR assure maintenance of the

same margin of safety as presently exists for

the prevention of onset of transition boiling.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: May 30,

1996

Description of amendment request:

The proposed amendment would revise

Anticipated Transient Without Scram

(ATWS) Recirculation Pump Trip

Reactor Pressure - High setpoint when

either zero or one Safety Relief Valves

are out-of service.

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:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

A change in the ATWS Recirculation Pump

Trip Reactor Pressure - High setpoint does

not affect initiation of any accident.

Operation in accordance with the revised

setpoints ensures the consequences of

previously analyzed accidents are not

changed.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

RPV [reactor pressure vessel] pressure

following an ATWS with MSIV [main steam

isolation valve] closure event (worst case

transient for RPV pressurization) remains

within acceptable limits with the revised

setpoint. Therefore changing the setpoint will

not lead to a new type of accident.

3. Involve a significant reduction in a

margin of safety because:

The analyses performed to determine the

revised ATWS Recirculation Pump Trip

Reactor Pressure - High setpoint assure

maintenance of the same margin of safety as

presently exists for limiting RPV pressure

following an ATWS with MSIV closure

(limiting transient).

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: May 30,

1996

Description of amendment request:

The proposed amendment would

eliminate selected response time testing

requirements. The affected Technical

Specifications (TS) are TS 4.1.A,

‘‘Surveillance Requirements, Reactor

Protection System,’’ and TS 4.2.A,

‘‘Surveillance Requirements,

Instrumentation, Primary Containment

Isolation Functions.’’

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:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

The purpose of the proposed TS change is

to eliminate response time testing

requirements for selected sensors in the RPS

[reactor protection system] and Primary

Containment Isolation System. The BWROG

[Boiling Water Reactor Owners Group] has

completed an evaluation which demonstrates

that response time testing is redundant to the

other TS required testing. These other tests
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34897

in conjunction with actions taken in response

to NRC Bulletin 90-01, ‘‘Loss of Fill-Oil in

Transmitters

Manufactured by Rosemount,’’ and

Supplement 1 to Bulletin 90-01, are sufficient

to identify failure modes or degradation in

instrument response time and ensure

operation of the associated systems within

acceptable limits. Furthermore, failure modes

detected by response time testing are

detectable by other TS required testing. This

evaluation was documented in Reference 1

[See application dated May 30, 1996]. NYPA

[New York Power Authority] has confirmed

the applicability of this evaluation to the

FitzPatrick Plant. In addition, NYPA will

complete the actions identified in the NRC

staff’s safety evaluation of NEDO-32291-A.

Because of the continued application of

other existing TS required tests such as

channel calibrations, channel checks,

channel functional tests, and logic system

functional tests, the response time of these

systems will be maintained within the

acceptance limits assumed in plant safety

analyses and required for successful

mitigation of an initiating event. The

proposed changes do not affect the capability

of the associated systems to perform their

intended function within their required

response time, nor do the proposed changes

themselves affect the operation of any

equipment. Therefore, the proposed change

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated because:

The proposed changes do not affect the

ability of the systems to perform their

intended function within the acceptance

limits assumed in plant safety analyses and

required for successful mitigation of an

initiating event. No new failure modes are

introduced by the changes. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in the

margin of safety.

The current TS required response time test

limits are based on the maximum allowable

values assumed in the plant safety analyses.

These analyses conservatively establish the

margin of safety. As described above, the

proposed changes do not affect the capability

of the associated systems to perform their

intended function within the allowed

response time used as the basis for the plant

safety analysis. Plant and system response to

an initiating event will remain in compliance

within the assumptions of the safety

analyses, and therefore the margin of safety

is not affected.

Further, although not explicitly evaluated,

the proposed changes will provide an

improvement to plant safety and operation by

reducing the time safety systems are

unavailable, reducing safety systems

actuations, reducing plant shutdown risk,

limiting radiation exposure to plant

personnel, and eliminating the diversion of

key personnel to conduct unnecessary

testing. Therefore, the overall effect of the

changes should increase the margin the

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: March 6,

1996, as supplemented by letter dated

May 30, 1996

Description of amendment request:

The proposed change to Hope Creek

Technical Specification (TS) 3.8.1,

‘‘A.C. Sources - Operating’’, would

decrease the minimum fuel oil storage

capacity of the Emergency Diesel

Generator Fuel Oil Storage Tanks, from

48,800 to 44,800 gallons. In addition,

footnote ** is deleted from TS

3.8.1.1.b.2. The proposed change would

also add an Action Statement to address

remedial action when a fuel oil transfer

pump becomes inoperable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

TANK LEVEL

Amendment 59 provides an allowance for

transferring fuel oil from a pair of storage

tanks associated with an inoperable

[Emergency Diesel Generator] EDG to another

pair of storage tanks in order to demonstrate

compliance with PSE&G’s commitment to

Regulatory Guide 1.137. The proposed

change is consistent with that transfer

strategy and extends this allowance to

include using fuel oil in operable EDG

storage tanks in order to reduce the amount

of stored fuel oil. Transfer from operable EDG

storage tanks is, actually, less complex than

transferring from an inoperable EDG storage

tank since power to the transfer pumps

would be available.

The low level alarm setpoint is the only

physical change to be made. No change is

being made to the EDGs, to the fuel oil

storage tanks, or to the fuel oil transfer

system and since EDG fuel oil supply is

associated with mitigating the consequences

of an accident, there is no change in the

probability of any accident analyzed in the

[Updated Final Safety Analysis Report]

UFSAR.

Since the proposed change still ensures the

minimum fuel oil storage capacity meets the

existing licensing basis and since off-site

replacement oil is expected to be available

within 60 hours there is no change in the

consequences of an accident previously

evaluated.

TRANSFER PUMP ACTION STATEMENT

Since no change is being made to the

EDGs, to the fuel oil storage tanks or to the

fuel oil transfer system, and since EDG fuel

oil supply is associated with mitigating the

consequences of an accident, there is no

change in the probability of any accident

analyzed in the UFSAR.

The proposed change provides

compensatory action in the event a single

fuel oil transfer pump is inoperable without

having to immediately declare the EDG

inoperable. The change ensures the affected

EDG remains fully capable of functioning as

assumed in the safety analyses, therefore,

there is no significant impact on the

consequences of an accident previously

evaluated.

Therefore, the proposed changes will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

TANK LEVEL AND TRANSFER PUMP

ACTION STATEMENT

The proposed changes will result in a

setpoint change to the low level alarm. No

other physical changes to the EDGs, to the

fuel oil storage tanks, or to the fuel oil

transfer system will result from the proposed

changes. Operation including the proposed

changes will not impair the diesel generators

from performing as provided in the design

basis. In addition, EDG fuel oil supply is

associated with mitigating accident

consequences, not accident prevention.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Will not involve significant reduction in

a margin of safety.

TANK LEVEL

The margin of safety is provided by the onsite storage of an adequate supply of diesel

fuel oil to ensure uninterrupted EDG

operation for seven days. Although the

proposed change may result in a reduction of

stored fuel oil, the new minimum continues

to provide for an on-site seven day supply of

diesel fuel oil.

TRANSFER PUMP ACTION STATEMENT

The margin of safety is provided by the

ability of the fuel oil transfer pumps to

supply an adequate flow of the stored fuel to

each EDG day tank. The proposed change

continues to provide 100% capacity to the

EDG day tank for a minimum of three days

with no operator action. With the proposed

action, adequate transfer capability is
34898 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

provided for a minimum of seven days fuel

oil supply at which time refilling of the tanks

would provide an indefinite supply. With

both transfer pumps on a single EDG

inoperable, the remaining three EDGs would

provide adequate power for safe shutdown.

Transfer of fuel oil from the storage tanks

with inoperable transfer pumps can still be

effected using temporary hoses.

Since the proposed changes do not involve

the addition of plant equipment, are

consistent with the intent of the existing

Technical Specifications, are consistent with

allowances for fuel oil transfers approved in

Amendment 59, meets the intent of

Regulatory Guide 1.137, and are consistent

with the design basis of the diesel generators

and the accident analysis, no action proposed

by this request will occur that will 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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: May 10,

1996

Description of amendment request:

The proposed amendments would

change Technical Specification

Sections, 1.0, 2.0, 3/4 1.0, 3/4 2.0, 5.0

and 6.0. These changes support the

Margin Recovery Program (MRP) and

support increased steam generator tube

plugging, improved fuel reliability,

reduced fuel costs, longer fuel cycles,

reduced spent fuel storage, and

enhanced reactor safety. These changes

incorporate the results of the revised

safety analyses (margin recovery) and

the establishment of a Core Operating

Limits Report.

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 accidents potentially affected by the

parameters and assumptions associated with

the MRP have been evaluated/ analyzed and

all design standards and applicable safety

criteria are met. The consideration of these

changes does not result in a situation where

the design, material, or construction

standards that were applicable prior to the

change have been altered. Therefore, the

changes occurring with the MRP will not

result in any additional challenges to plant

equipment that could increase the probability

of any previously evaluated accident.

The changes associated with the MRP do

not affect plant systems such that their

function in the control of radiological

consequences is adversely affected. The

safety evaluation documents that the design

standards and applicable safety criteria limits

continue to be met and therefore fission

barrier integrity is not challenged. The MRP

changes have been shown not to adversely

affect the response of the plant to postulated

accident scenarios. In all cases, the

calculated doses are within the regulatory

criteria and therefore do not constitute an

increase in consequences. These changes

will, therefore, not affect the mitigation of the

radiological consequences of any accident

described in the Updated Final Safety

Analysis Report (UFSAR).

Based on the above, it is concluded that the

probability or consequences of an accident

previously evaluated is not significantly

increased by the proposed changes.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The possibility for a new or difference[t]

type of accident from any accident

previously evaluated is not created since the

changes associated with the MRP do not

result in a change to the design basis of any

plant component or system. The evaluation

of the effects of the MRP changes shows that

all design standards and applicable safety

criteria limits are met. These changes

therefore do not cause the initiation of a new

accident nor create any new failure

mechanisms. Component integrity is not

challenged. The changes do not result in any

event previously deemed incredible being

made credible. The MRP changes will not

result in more adverse conditions and will

not result in any increase in the challenges

to safety systems.

Therefore, the consideration of the MRP as

described in the safety evaluation does not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety is maintained by

assuring compliance with acceptance limits

reviewed and approved by the NRC. Since all

of the appropriate acceptance criteria for the

various analyses and evaluations have been

met, by definition there has not been a

reduction in any margin of safety.

Therefore, the margin of safety as defined

in the Bases to the Salem Unit 1 and 2

Technical Specifications has not been

significantly reduced.

Based on the above, PSE&G has determined

that the proposed changes do not involve a

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

Authority, Docket No. 50-395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: April 16,

1996

Description of amendment request:

The proposed amendment would revise

the Virgil C. Summer Nuclear Station,

Unit 1 (VCSNS), Technical

Specifications (TS) to implement the

amended regulation to 10 CFR Part 50,

Appendix J, Option B (new rule), to

provide a performance-based option for

leakage-rate testing of containment. The

proposed amendment will revise the

VCSNS TS 3/4.6 ‘‘Containment

Systems,’’ TS Bases 3/4.6, and TS 6.8

‘‘Administrative Controls - Programs

and Procedures,’’ to adopt the

implementation requirements of 10 CFR

Part 50, Appendix J, Option B. The

proposed amendment utilizes the

guidelines (guidelines) provided in

‘‘Option B’’ of Regulatory Guide (RG)

1.163 ‘‘Performance-Based Containment

Leak-Test Program, September 1995,’’

and NEI 94-01, ‘‘Industry Guideline for

Implementing Performance-Based

Option of 10 CFR 50, Appendix J, July

26, 1995.’’ The licensee has stated that

the proposed amendment is within

these prescribed guidelines and does

not propose any deviations to the

established methods which would

impact already approved analyses/

justifications and established review

process.

The proposed change will remove the

prescriptive TS requirements for the

performance of containment leakage

testing and allow leakage testing to be

conducted as determined appropriate

through the performance-based or riskbased alternatives described in the

VCSNS Containment Leakage Rate

Testing Program developed in

accordance with RG 1.163 and NEI 94-

01. Since the requirements of Appendix

J to 10 CFR Part 50 will continue to
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34899

apply, the type of testing will not

change. The proposed request does not

modify any plant equipment or systems.

The requirements of Appendix J will

continue to govern the type of test,

testing methodology, and acceptance

criteria for Type A, B, and C testing. The

performance-based testing of Option B

eliminates or modifies prescriptive

regulatory requirements for which the

burden is marginal to safety for which

the reviews and analyses have been

presented in NUREG-1493,

‘‘Performance-Based Containment LeakTest Program, Final Report, September

1995.’’

Earlier leakage testing performed at

VCSNS has demonstrated low overall

containment leakage and supports the

implementation of Option B.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The probability or consequences of an

accident previously evaluated is not

significantly increased.

There is no increase in the probability of

an accident since there is no work that would

affect containment integrity. The testing of

containment isolation valves (CIVs) and other

containment penetration sealing devices is

not postulated as an accident precursor or

initiating event.

Type A testing is capable of determining

the total leakage from both local leakage

paths and gross containment leakage paths.

Our Type B and C testing has consistently

provided accurate leakage rates for valves

and penetrations.

Administrative controls govern

maintenance and testing such that there is

very low probability that unacceptable

maintenance or alignments can occur. Prior

to and following maintenance on CIVs and

penetrations, a local leak rate test (LLRT) is

required to be performed. As a result, Type

A testing is not required to accurately

quantify the leakage through containment

penetrations.

Any specific exemptions to the

requirements of Appendix J will require

approval by the NRC before implementation.

Therefore, this proposed change does not

involve a significant increase in the

possibility or consequences of an accident

previously evaluated.

2. The possibility of an accident or a

malfunction of a different type than any

previously evaluated is not created.

The proposed request does not involve any

physical changes to the plant, affect the

operation of the plant, or change testing

methods or acceptance criteria. The history

of containment testing verifies that

containment integrity has been maintained.

The frequency changes allowed by

implementation of Option B will not

significantly decrease the level of confidence

in the ability of the reactor building to limit

offsite doses to allowable values. No accident

or malfunction can be the result of the

allowed changes to test schedule or

frequency.

Since the proposed request will not

directly impact equipment, procedures or

operations, the changes will not create the

possibility of any new or different kind of

accident from any previously evaluated.

3. The margin of safety has not been

significantly reduced.

The reason for performing containment

leakage rate testing is to assure that the

leakage paths are identified, and that any

accident release will be restricted to those

paths assumed in the safety analysis. The

purpose for the schedule is to assure that

containment integrity is verified on a

periodic basis.

Implementation of Option B to provide

flexibility in the scheduled requirements

does not mean that containment integrity

will be compromised. The historical leakage

rate test results for VCSNS and for the

nuclear industry support extension of testing

frequencies and demonstrate that structural

integrity has been maintained.

Therefore, the margin of safety has not

been significantly reduced.

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: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, SC 29218

NRC Project Director: Eugene V.

Imbro

Southern Nuclear Operating Company,

Inc., Docket No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: April 22,

1996

Description of amendment request:

The amendment would revise the

Technical Specifications to implement

the L* Tubesheet Region Plugging

Criterion, which would allow a steam

generator tube to remain in service with

bands of axial degradation in the

tubesheet region provided sufficient

non-degraded tubing remains to satisfy

regulatory guidance concerning

structural and leakage integrity.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the Farley Nuclear Plant

Unit 2 steam generators in accordance with

the proposed license amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The supporting technical evaluations of the

subject criteria demonstrate that the presence

of the tubesheet enhances the tube integrity

in the region of the hardroll by precluding

tube deformation beyond its initial expanded

outside diameter. The resistance to both tube

rupture and tube collapse is strengthened by

the presence of the tubesheet in that region.

The result of the hardroll of the tube into the

tubesheet is an interference fit between the

tube and the tubesheet. Tube rupture

[cannot] occur because the contact between

the tube and tubesheet does not permit

sufficient movement of tube material. In a

similar manner, the tubesheet does not

permit sufficient movement of tube material

to permit buckling collapse of the tube

during postulated LOCA [loss-of-coolant

accident] loadings.

The type of degradation for which the L*

criterion has been developed (cracking with

an axial or near axial orientation) has been

found not to significantly reduce the axial

strength of a tube. An evaluation including

analysis and testing has been done to

determine the strength reduction for axial

loads with simulated axial and near axial

cracks. This evaluation provides the basis for

the acceptance criteria for tube degradation

subject to the L* criterion.

The SRE [sound roll expansion] L* length

is sufficient to preclude significant leakage

from tube degradation located below the L*

length. The existing Technical Specification

leak rate requirements and accident analysis

assumptions remain unchanged in the

unlikely event that significant leakage from

this region does occur. Any leakage from the

tube within the tubesheet at any elevation in

the tubesheet is fully bounded by the existing

steam generator tube rupture analysis

included in the Farley Nuclear Plant Final

Safety Analysis Report. A conservative

leakage allowance for each L* tube is

provided to determine the impact of L*

criterion upon offsite doses in the event of a

postulated double ended guillotine break of

the main steam line outside of containment,

but upstream of the main steam line isolation

valves. Since Farley Unit 2 has implemented

the Interim Plugging Criteria (IPC) for ODSCC

at the tube support plates, projected steam

line break (SLB) leakage at the end of the

next successive operating cycle must be

evaluated. Per Generic Letter 95-05, plants

implementing the IPC can utilize SLB leakage

limits higher than the originally assumed 1.0

gpm primary to secondary leakage value

provided an analysis of offsite doses

consistent with the Standard Review Plan

methodology is performed. This analysis

performed for the Farley Unit 2 plant

indicates that primary to secondary leakage

of 11.2 gpm in the faulted loop (0.1 gpm in

the intact loops) will result in offsite doses

at the site boundary of less than 10% of the

10 CFR [Part] 100 guidelines. The total

projected SLB leakage from all leakage

sources must remain below this value. Per

attachment 4 addressing the L* methodology,
34900 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

the number of tube ends to which L*

criterion can be applied is limited to 600 per

steam generator. Using a bounding SLB

leakage allowance per L* tube, the SLB

leakage component from 600 L* tube ends

will be less than 0.33 gpm in the faulted

loop. The proposed alternate plugging

criterion does not adversely impact any other

previously evaluated design basis accident.

As the current Unit 2 IPC SLB leakage has

been calculated to be less than 2 gpm in the

faulted loop, [an] SLB leakage margin of over

9 gpm is provided for this cycle.

As noted above, tube rupture and pullout

is not expected for tubes using the L*

criterion. In addition to the L* length, a

minimum length of SRE below the identified

degradation must be established. The

aggregate L* distance of SRE provides the

structural integrity to prevent tube pullout.

Conservatively, it is assumed that the

degraded band length does not provide any

support in resisting tube pullout.

Therefore SNC [Southern Nuclear

Operating Company, Inc.] concludes that

Operation of the Farley Nuclear Plant Unit 2

steam generators in accordance with the

proposed license amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Implementation of the proposed L*

criterion does not introduce any significant

changes to the plant design basis. Use of the

criterion does not provide a mechanism to

result in an accident initiated outside of the

region of the tubesheet expansion. The

structural integrity of L* tubes will be

maintained during all plant conditions. Any

hypothetical accident as a result of any tube

degradation in the expanded portion of the

tube would be bounded by the existing tube

rupture accident analysis. If it is postulated

that a circumferential separation of an L*

tube were to occur below the PLRL [pullout

load reaction length], tube structural and

leakage integrity will be maintained during

all plant conditions. Verification of the L*

distance of non-degraded tube roll expansion

prevents the postulated separated tube from

lifting out of the tubesheet during all plant

conditions. Verification of the L* criterion

prevents tube displacement of any

magnitude, and therefore, postulated axial

cracks existing a minimum of 0.5 inch from

either the bottom of the roll transition or top

of tubesheet, whichever is lower, from

migrating out of the tubesheet.

Therefore, SNC concludes that the

proposed license amendment does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed license amendment does

not involve a significant reduction in a

margin of safety.

The use of the L* criterion has been

concluded to maintain the integrity of the

tube bundle commensurate with the

requirements of draft Regulatory Guide 1.121

under normal and postulated accident

conditions. The safety factors used in the

verification of the strength of the degraded

tube are consistent with the safety factors in

the ASME [American Society of Mechanical

Engineers] Boiler and Pressure Vessel Code

used in steam generator design. The L*

length has been verified by testing to be

greater than the length of roll expansion

required to preclude significant leakage

during normal and postulated accident

conditions. The leak testing acceptance

criteria are based on the primary to

secondary leakage limit in the Technical

Specifications and the leakage assumptions

used in the FSAR accident analyses. The L*

distance provides for structural integrity

during all plant conditions.

Implementation of the L* criterion will

decrease the number of tubes which must be

taken out of service with tube plugs or

repaired with sleeves. Both plugs and sleeves

reduce the RCS [reactor coolant system] flow

margin, thus implementation of the L*

criterion will maintain the margin of flow

that would otherwise be reduced in the event

of increased plugging or sleeving.

Therefore, SNC, concludes based on the

above, it is concluded that the proposed

change does not result in a significant

reduction in a loss of margin with respect to

plant safety as defined in the Final Safety

Analysis Report [FSAR] or the bases of the

FNP [Farley Nuclear Plant] technical

specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: May 29,

1996

Description of amendment request:

The application requests staff review

and approval of a modification to the

facility, as described in the safety

analysis report, that involves an

unreviewed safety question. The

modification will reduce the single

failure trip potential for the main

feedwater control and bypass valves.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

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 Callaway safety analysis assumes the

MFC&BVs [main feedwater control and

bypass valves] close during certain events in

order to terminate fluid inventory addition to

faulted steam generators and thereby

preclude the diversion of auxiliary feedwater

to the main feedwater system. This feature is

necessary because each feedwater line at

Callaway is equipped with only one MFIV

[main feedwater isolation valve]. It should be

noted that the safety analysis simply requires

the valves to close and does not prescribe a

mechanism for accomplishing that action.

The following are accidents that credit

feedwater isolation or AFW [auxiliary

feedwater] addition. There is no impact by

the proposed modification on the

consequences of each accident.

•Feedwater System Malfunctions That

Result In An Increase In Feedwater Flow

•Inadvertent Opening Of A Steam

Generator Relief or Safety Valve

•Steam System Piping Failure

•Loss of Nonemergency AC Power to the

Station Auxiliaries

•Loss of Normal Feedwater Flow

•Feedwater System Pipe Break

•Decrease in Reactor Coolant Inventory

The modification will not change the

radiological consequences of FSAR [final

safety analysis report] Chapter 15 accidents

because the feedwater isolation function (and

NSSS [nuclear steam supply system] break

response) has not changed. Therefore, there

will be no increase in the consequences of an

accident evaluated previously in the FSAR.

An analysis was performed to quantify the

impact of the proposed modification on the

probability of MFCV [main feedwater control

valve] failure (closure) during normal plant

operation. Comparison of this failure

probability for the existing design (1.20E-1

per year) versus the proposed design (6.99E2 per year) indicates that the percentage

reduction in the system failure probability at

power is 41.75%. Thus, the proposed design

results in a reduction in the probability of

inadvertent MFCV failures at power and

hence, a reduction in the probability of a

reactor trip and subsequent challenges to

other safety systems.

While this modification reduces the

probability of a reactor trip, it slightly

increases the unavailability of the feedwater

isolation function. This is because the

original design required actuation of only one

FWIS [feedwater isolation system] train to

close the MFC&BVs, whereas the new design

requires actuation of both trains. The impact

of the modification on the probability of

incurring a feedwater isolation failure was

therefore quantified, utilizing PRA

[probabilistic risk assessment] techniques.

Fault trees were developed for both the new

and existing designs. Failure probabilities for

each event were then obtained from the IPE

[individual plant examination] and utilized

to calculate failure probabilities for the

feedwater isolation safety function. This

calculation considered hardware failures
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34901

only, i.e., failure of an MFIV to close after

receiving an actuation signal. The failure

probability of feedwater isolation, based on

the proposed design, was determined to be

6.1E-5 per demand (1 event every 16,400

demands). The existing design was found to

have a failure probability of 2.8E-5 per

demand (1 event every 35,700 demands).

Therefore, this modification will not

significantly increase the probability or

consequences of an accident evaluated

previously in the FSAR.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The modification maintains the present deenergize-to-actuate configuration of the

MFC&BV trip solenoid valves.

Thus, the proposed modification does not

create the possibility of an accident of a

different type than any previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Credit is taken in the accident analyses for

the MFIVs to close on demand for feedwater

isolation. Because of this, the MFIVs have

been incorporated into the Callaway

Technical Specifications. Action Statements

and surveillance requirements have been

developed to assure the availability of the

valves when needed.

The MFC&BVs are not addressed by any of

the Callaway Technical Specifications or

their bases. Therefore, this modification will

not involve a significant reduction in the

margin of safety as defined in the basis for

any technical specification.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: William H.

Bateman

Wisconsin Electric Power Company,

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

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: May 29,

1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

15.4.4, ‘‘Containment Tests,’’ to

incorporate the provisions of 10 CFR

Part 50, Appendix J, ‘‘Primary Reactor

Containment Leakage Testing for WaterCooled Power Reactors,’’ Option B.

Revisions would also be made to TS

Sections 15.1, ‘‘Definitions,’’ 15.3.6,

‘‘Containment System,’’ and 15.6,

‘‘Administrative Controls,’’ to support

the proposed changes to Section 15.4.4.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The proposed change does not involve a

change to structures, systems, or components

which would affect the probability or

consequences of an accident previously

evaluated in the PBNP [Point Beach Nuclear

Plant] Final Safety Analyses Report (FSAR).

Furthermore, containment leakage rate

testing is not an initiator of any accident. The

proposed change simply provides a

mechanism within the Technical

Specifications for implementing a

performance-based method of determining

the frequency for leakage rate testing which

has been approved by the NRC. The proposed

change does not affect reactor operations or

accident analysis and has no significant

radiological consequences. Therefore, this

change will not create a significant increase

in the probability or consequences of an

accident previously evaluated.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not involve a

change to the plant design or operation. As

a result, the proposed change does not affect

any of the parameters or conditions that

contribute to initiation of any accidents. This

change involves a potential 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 will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The proposed change potentially affects

only the frequency of Type A, B, and C

testing. Except for the method of defining test

frequency, the methods for performing the

actual tests are not changed. The proposed

change is based on NRC accepted provisions

and maintains necessary levels of system and

component reliability affecting containment

integrity. Evaluation of the performancebased approach to leakage rate testing, as

documented in NUREG-1493, concludes that

the impact on public health and safety due

to revised testing intervals is negligible.

Furthermore, the proposed change will not

reduce the availability of systems associated

with containment integrity when they are

required to mitigate accident conditions.

Therefore, the proposed change will not

create 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: June 4,

1996

Description of amendment request:

The proposed amendment would revise

the Kewaunee Nuclear Power Plant

Technical Specifications (TS) by

reducing the surveillance test

frequencies for the radiation monitoring

system (Table TS 4.1-1) and the control

rods (Table TS 4.1-3) in accordance with

the guidance of Generic Letter 93-05,

‘‘Line-Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation,’’ dated September 27, 1993.

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:

Table TS 4.1-1, ‘‘Minimum Frequencies for

Checks, Calibrations and Test of Instrument

Channels,’’ Item 19

The proposed changes were reviewed in

accordance with the provisions of 10 CFR

50.92 to determine that no significant

hazards exist. The proposed changes will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The radiation monitors are not accident

initiators; therefore, they cannot increase the

probability of an accident occurring. The

reliability of the radiation monitors is not

expected to decrease due to the decreased

surveillance frequency; therefore, this change

does not increase the consequences of an

accident.
34902 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

The addition of comment (a) to the Check,

Calibrate, and Test columns is merely a

clarification of the existing information in the

table and does not change the intent of the

Technical Specifications.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change revises only the

testing frequency and does not revise the test

method or operational performance of the

radiation monitors. The radiation monitors

are not accident initiators; therefore, they

cannot create a new or different kind of

accident.

3. Involve a significant reduction in the

margin of safety.

Quarterly testing of the radiation

monitoring system channels will continue to

verify operability of the monitors. Decreasing

the test surveillance frequency is not

expected to decrease the reliability of the

radiation monitors. This change is acceptable

in accordance with Generic Letter 93-05 and

NUREG-1366, ‘‘Improvements to Technical

Specifications Surveillance Requirements.’’

Table TS 4.1-3, ‘‘Minimum Frequencies for

Equipment Tests,’’ Item 1

The proposed change in test frequency for

control rod exercising was reviewed in

accordance with the provisions of 10 CFR

50.92 to determine that no significant

hazards exist. It has been determined that the

proposed change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change revises only the

testing frequency for control rod exercising.

The control rod exercise surveillance

procedure will continue to be conducted, on

a quarterly basis, to ensure that the

equipment remains operable. The reduced

frequency of control rod exercising reduces

the probability of an inadvertent reactor trip

occurring during testing due to a dropped

control rod. Surveillance procedure SP 49-

075 is conducted to verify rod movement. In

accordance with NUREG-1366, the frequency

of a stuck control rod occurring is very low.

This condition is most often discovered

during reactor startup or during low power

physics testing. The reduction in control rod

exercising is, therefore, considered

acceptable and is not expected to affect the

probability of a stuck control rod occurring.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change revises only the

testing frequency and does not revise the test

method or the design of the control rod

system. Therefore, a new or different kind of

accident will not be created by this change.

3. Involve a significant reduction in the

margin of safety.

Quarterly control rod exercising will

continue to verify movement of the control

rods. No adverse consequences are expected

to occur due to decreasing the test frequency.

This change is acceptable in accordance with

Generic Letter 93-05 and NUREG-1366.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Attorney for licensee: Bradley D.

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

Box 1497, Madison, Wisconsin 53701-

1497

NRC Project Director: Gail H. Marcus

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: June 10,

1996

Description of amendment request:

The proposed amendment would revise

Technical Specification 4.2.b, ‘‘Steam

Generator Tubes,’’ and its associated

basis, by allowing the use of

Westinghouse laser-welded sleeves to

repair defective steam generator tubes.

A description of the sleeving repair

process and supporting technical

justification are contained in WCAP13088, Revision 3, ‘‘Westinghouse

Series 44 and 51 Steam Generator

Generic Sleeving Report.’’ WCAP-13088,

and a non-proprietary version (WCAP13089), were submitted to the Nuclear

Regulatory Commission on April 13,

1995, to support a similar TS

amendment request for the DC Cook

Nuclear Power Plant, Unit 1.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the KNPP [Kewaunee

Nuclear Power Plant] in accordance with the

proposed license amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The LWS [laser-welded sleeve]

configuration has been designed and

analyzed in accordance with the

requirements of the ASME [American Society

of Mechanical Engineers] Code. Fatigue and

stress analyses of the sleeved tube assemblies

produced acceptable results; i.e., the applied

stresses and fatigue usage for the sleeve and

weld are bounded by the limits established

in the ASME Code. ASME Code minimum

material property values are used for the

structural and plugging limit analysis.

Ultrasonic inspection is used to verify that

minimum weld fusion zone thicknesses are

produced. Mechanical testing of 7/8’’

tubesheet sleeves installed in roll expanded

tubes has shown that the individual joint

structural strength of Alloy 690 LWSs

provides margin to acceptance limits. These

acceptance limits bound the most limiting

loadings (3 times normal operating pressure

differential) recommended by RG [Regulatory

Guide] 1.121. Therefore, each individual

joint provides for structural integrity

exceeding RG recommendations. A

hypothetical loss of integrity of one of the

joints will not result in a loss of structural

integrity for the sleeve. Leakage testing for 3/

4’’ and 7/8’’ full length tubesheet sleeves has

demonstrated that unacceptable levels of

primary-to-secondary leakage are not

expected during all plant conditions for nonwelded tubesheet sleeve lower joints. The

welded joint produces a hermetic seal, and

therefore will not leak under any plant

conditions. Laser welded sleeves will not

contribute to the current SLB [steam-line

break] primary-to-secondary leakage limit of

34 gpm in the faulted loop. The 34 gpm

leakage limit was calculated in accordance

with the standard review plan methodology

to support implementation of the voltagebased repair criteria for tube support plate

intersections.

The sleeve minimum acceptable wall

thickness (used for developing the depth

based plugging limit for the sleeve) is

determined using the guidance of RG 1.121

and the pressure stress equation of Section III

of the ASME Code. With respect to the design

of the sleeve for KNPP, the limiting

requirement of the RG which applies to part

throughwall degradation is that the minimum

acceptable wall must maintain a factor of

safety consistent with the analysis conditions

as defined by the ASME Code. A bounding

set of design and transient loading input

conditions was used for the minimum wall

thickness evaluation in the generic

evaluation. Evaluation of the minimum

acceptable wall thickness for normal, upset

and postulated accident condition loading

per the ASME Code indicates the limiting

condition is established for the normal

operating conditions, and the minimum

acceptable wall thickness for this case

bounds the upset and faulted condition

values.

According to RG recommendations, an

allowance for non-destructive evaluation

(NDE) uncertainty and operational growth of

existing tube wall degradation indications

within the sleeve must be accounted for

when determining the sleeve plugging limit.

A conservative tube wall degradation growth

rate per cycle and an NDE uncertainty has

been assumed for determining the sleeve TS

plugging limit. The sleeve wall degradation

extent determined by NDE, which would

require plugging sleeved tubes, is developed

using the guidance of RG 1.121 and is

defined in WCAP-13088 [non-proprietary

WCAP-13089] to be 25% throughwall

(plugging limit = 100% - structural limit +

NDE uncertainty + growth) for KNPP.

The hypothetical consequences of failure

of the sleeve joint would be bounded by the

current SG [steam generator] tube rupture

analysis included in the KNPP Updated

Safety Analysis Report. Due to the slight

reduction in diameter caused by the sleeve

wall thickness, primary coolant release rates

would be slightly less than assumed for the

SG tube rupture analysis (depending on

break location), and therefore, would result
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34903

in lower total primary fluid mass release to

the secondary system.

The proposed TS change to use Alloy 690

LWSs does not adversely impact any other

previously evaluated design basis accidents

or the results of LOCA [loss of coolant

accident] and non-LOCA accident analyses

for the current TS minimum reactor coolant

system flow rate. The results of the analyses

and testing, as well as plant operating

experience, demonstrates that the sleeve

assembly is an acceptable means of

maintaining tube integrity. Plugging limit

criteria are established using the guidance of

RG 1.121. Furthermore, per RG 1.83

recommendations, the sleeved tube will be

monitored through periodic inspections with

present NDE techniques. These measures

demonstrate that installation of sleeves

spanning degraded areas of the tube will

restore the tube to a condition consistent

with its original design basis.

Corrosion testing of free span LWS joint

has indicated that the corrosion resistance

(relative to roll transitions) can be increased

by greater than a factor of ten with the

application of a PWHT [post weld heat

treatment] step. Estimations of joint

susceptibility based on expected far field

stresses after heat treatment using the

expected original tube-to-tubesheet hydraulic

expansion transition residual stresses and

actual time to crack in these transitions at

KNPP indicate that LWS joint lifetime should

exceed the current plant license. Consistent

with other license amendments addressing

LWS, all free span laser welds will receive

a PWHT; therefore, rapid corrosion

degradation of the free span joint is not

expected. Recently performed corrosion

testing of LWS joints in locked tube

conditions indicates that with PWHT the

stress corrosion cracking resistance and

initiation potential in the parent tube weld

region is greatly enhanced. Similar test

results and conclusions would be expected

for KNPP. The Model 51 SG tube span

between the top of the tubesheet and the first

support plate is such that even lower PWHT

residual stresses would be expected. Also,

the weld placement within the hydraulically

expanded area and sleeve installation

sequence have been optimized to provide for

some level of heat treatment at the upper

transition above the weld and lower far field

residual stress levels. While no parent tube

degradation has been detected at this

elevation, or any other elevation in a laser

welded sleeve assembly, the relocation of the

weld serves to provide further resistance to

PWSCC [primary water stress corrosion

cracking] at this elevation. The suggested

target PWHT temperature has also been

optimized in that this temperature provides

for adequate PWHT while maintaining the

parent tube far field stresses.

Approximately 19,500 LWSs have been

installed in the U.S. Of this number, over 300

which have up to 3 cycles of operation were

inspected in 1995 using the CECCO-5 probe.

No degradation of the sleeves or the parent

tube was detected. Operating experience in

Europe has shown good performance of the

LWS joint for up to 5 cycles of operation. In

1994, approximately 11,200 LWSs were

installed in the Doel-4 Plant. After one year

of operation, all in-service sleeves were

inspected using the +point probe. No service

induced corrosion was detected. In 1995,

approximately 18,600 LWSs were installed in

two different U.S. plants. Due to their limited

operational time, these sleeves have not been

inspected.

Conformance of sleeve design with the

applicable sections of the ASME Code and

results of the leakage and mechanical tests

support the conclusion that installation of

LWSs will not increase the probability or

consequences of an accident previously

evaluated.

2. The proposed license amendment

request does not create the possibility of a

new or different kind of accident from an

accident previously evaluated.

Installation of LWSs will not introduce

significant or adverse changes to the plant

design basis and does not represent a

potential to affect any other plant

component. Stress and fatigue analysis of the

repair has shown that the ASME Code and

RG 1.121 criteria are not exceeded.

Installation of LWSs maintains overall tube

bundle structural and leakage integrity at a

level consistent to that of the originally

supplied tubing during all plant conditions;

stresses are bounded by the Code and the

tubing is leaktight. Sleeving of tubes does not

provide a mechanism resulting in an accident

outside of the area affected by the sleeves.

Any hypothetical accident as a result of

potential tube or sleeve degradation in the

repaired portion of the tube is bounded by

the existing tube rupture accident analysis.

Since the sleeve design does not affect any

component or location of the tube outside of

the immediate area repaired, in addition to

the fact that the installation of sleeves and

the impact on current plugging level analyses

is accounted for, the possibility that laser

welded sleeving creates a new or different

type of accident is not supported.

Installation of LWSs will reduce the

potential for primary-to-secondary leakage

during postulated steam line break while not

significantly impacting primary coolant flow

area in the event of a LOCA. By effectively

isolating degraded areas of the tube through

repair, the potential for steam line break

leakage is reduced.

3. The proposed license amendment does

not involve a significant reduction in the

margin of safety.

The LWS repair of degraded SG tubes as

identified in WCAP-13088 [non-proprietary

WCAP-13089] has been shown by analysis to

restore the integrity of the tube bundle

consistent with its original design basis

conditions; i.e., tube/sleeve operational and

faulted conditions stresses and cumulative

fatigue usage are bounded by the ASME Code

requirements and the repaired tubes are

leaktight. The safety factors used in the

design of sleeves for the repair of degraded

tubes are consistent with the safety factors in

the ASME Code used in SG design. The

design of the LWS lower joint for 7/8’’ tube

sleeves has been verified by testing to

sufficiently preclude leakage during normal

and postulated accident conditions. The

portions of the installed sleeve assembly

which represents the reactor coolant pressure

boundary will be monitored for the initiation

and progression of sleeve/tube wall

degradation, thus satisfying the requirements

of RG 1.83. The portion of the tube bridged

by the sleeve joints is effectively removed

from the pressure boundary, and the sleeve

then forms the new pressure boundary. The

areas of the sleeved tube assembly which

require inspection are defined in WCAP13088 [non-proprietary WCAP-13089]. Since

the installed sleeves represent a portion of

the pressure boundary, a baseline inspection

of these areas is required prior to operation

with sleeves installed.

The effect of sleeving on the design

transients and accident analyses has been

reviewed based on the installation of sleeves

up to the level of SG tube plugging

coincident with the minimum reactor coolant

flow rate. The installation of sleeves is

evaluated as the equivalent of some level of

SG tube plugging. This is based on

determining the minimum reactor coolant

flow for the LOCA evaluation. Information

provided in WCAP-13088 [non-proprietary

WCAP-13089] describes the method to

determine the flow equivalent for all

combinations of tubesheet and tube support

plate sleeves. Therefore, installation of LWSs

will not result in a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Attorney for licensee: Bradley D.

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

Box 1497, Madison, Wisconsin 53701-

1497

NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company,

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

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: June 4,

1996 (VPNPD-96-035)

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

15.2.3, ‘‘Limiting Safety System Settings

and Protective Instrumentation,’’ and

Section 15.5.3, ‘‘Design Features -

Reactor,’’ to incorporate changes

associated with the operation of Point

Beach Nuclear Plant (PBNP), Unit 2,

with replacement steam generators.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the

proposed Technical Specifications will not
34904 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The proposed changes do not involve a

change to structures, systems, or components

which would affect the probability or

consequences of an accident previously

evaluated in the PBNP Final Safety Analyses

Report (FSAR). The proposed setpoints

maintain the margin to safe operation of Unit

2 with the replacement steam generators. In

order to maintain one set of safety analyses

for both units, the analyses for operation of

Unit 2 with the replacement steam generators

were performed to encompass the operation

of Unit 1. Therefore, the proposed changes

apply to the operation of both units and

maintain the margin of safety for each. The

proposed change to the description of

nominal RCS [reactor coolant system] volume

is an administrative change and has no effect

on plant operation. Therefore, the probability

or consequences of a previously evaluated

accident are not significantly increased as a

result of these changes.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not involve a

change to the plant design. The proposed

setpoints maintain the margin to safe

operation of Unit 2 with the replacement

steam generators. In order to maintain one set

of safety analyses for both units, the analyses

for operation of Unit 2 with the replacement

steam generators were performed to

encompass the operation of Unit 1.

Therefore, the proposed changes apply to the

operation of both units and maintain the

margin of safety for each. These changes do

not affect any of the parameters or conditions

that contribute to initiation of any accidents.

The proposed change to the description of

nominal RCS volume is an administrative

change and has no effect on plant operation

or initiation of any accidents. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The proposed setpoints maintain the

margin to safe operation of Unit 2 with the

replacement steam generators. In order to

maintain one set of safety analyses for both

units, the analyses for operation of Unit 2

with replacement steam generators were

performed to encompass the operation of

Unit 1. Therefore, the proposed changes

apply to the operation of both units and

maintain the margin of safety for each. The

proposed change to the description of

nominal RCS volume is an administrative

change and has no effect on plant operation.

Therefore, the proposed changes will not

create 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company,

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

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: June 4,

1996 (VPNPD-96-036)

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

15.2.1, ‘‘Safety Limit, Reactor Core,’’

15.2.3, ‘‘Limiting Safety System

Settings, Protective Instrumentation,’’

and Section 15.3.1.G, ‘‘Operational

Limitations,’’ to maintain safety margin

for Unit 2 with replacement steam

generators.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The proposed change does not involve a

change to structures, systems, or components

which would affect the probability or

consequences of an accident previously

evaluated in the PBNP [Point Beach Nuclear

Plant] Final Safety Analyses Report (FSAR).

The proposed changes maintain the margin

to safe operation for Unit 2 with the

replacement steam generators. In order to

maintain one set of safety analyses for both

units, the analyses for operation of Unit 2

with the replacement steam generators were

performed to encompass the operation of

Unit 1. Therefore, the proposed changes

apply to the operation of both units and

maintain the margin of safety for each. The

proposed changes do not change, degrade, or

preclude the prevention or mitigation of the

consequences of any accident described in

the FSAR. Therefore, the probability or

consequences of a previously evaluated

accident are not significantly increased as a

result of these changes.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not involve a

change to the plant design. The proposed

changes maintain the margin to safe

operation for Unit 2 with the replacement

steam generators. In order to maintain one set

of safety analyses for both units, the analyses

for operation of Unit 2 with the replacement

steam generators were performed to

encompass the operation of Unit 1.

Therefore, the proposed changes apply to the

operation of both units and maintain the

margin of safety for each. These changes do

not affect any of the parameters or conditions

that contribute to initiation of any accidents.

In addition, the safety functions of safetyrelated systems and components, which are

related to accident mitigation, have not been

altered. Therefore, the proposed changes will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The proposed changes maintain the margin

to safe operation for Unit 2 with the

replacement steam generators. In order to

maintain one set of safety analyses for both

units, the analyses for operation of Unit 2

with replacement steam generators were

performed to encompass the operation of

Unit 1. Therefore, the proposed changes

apply to the operation of both units and

maintain the margin of safety for each. The

proposed changes have no affect on the

availability, operability, or performance of

the safety-related systems and components

described in the Technical Specifications.

Therefore, the proposed changes will not

create 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

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
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34905

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.

Florida Power and Light Company, et

al., Docket No. 50-335, St. Lucie Plant,

Unit No. 1, St. Lucie County, Florida

Date of amendment request: June 1,

1996

Description of amendment request:

Revise Technical Specifications to

reflect reduced reactor coolant system

flows resulting from increased

percentage of plugged steam generator

tubes.

Date of publication of individual

notice in the Federal Register: June 7,

1996 (61 FR 29140)

Expiration date of individual notice:

June 24, 1996

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: May 31,

1996

Brief description of amendment

request: The amendments (1) revise the

Reactor Vessel Level Indication System

(RVLIS) Action Statements to facilitate

actions necessary for channel testing to

be performed in Mode 3, (2) revise the

Channel Calibration definition to better

account for temperature detector

channel calibration methodology, and

(3) delete a requirement to install a

jumper in the Auxiliary Feedwater

actuation logic since a design change

will result in the jumper function being

performed by a relay.

Date of publication of individual

notice in Federal Register: June 17, 1996

(61 FR 30641)

Expiration date of individual notice:

July 17, 1996

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

May 28, 1996, as supplemented by

letters dated May 31 and June 5, 1996

Brief description of amendments:

These amendments authorize the

licensee to revise applicable Updated

Final Safety Analysis Report sections to

reflect the installation of a variable flow

controller for the service water inlet

control valves for the containment air

coolers that is not within the current

licensing basis of Calvert Cliffs Nuclear

Power Plant Units No. 1 and No. 2.

These amendments are being issued

pursuant to the requirements of 10 CFR

50.59(c) because the review by

Baltimore Gas and Electric Company

identified the changes as an unreviewed

safety question. No changes to the

Technical Specifications are required by

these amendments.

Date of issuance: June 17, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 215 and 192

Facility Operating License Nos. DPR53 and DPR-69: The amendments

revised the Updated Final Safety

Analysis Report. Public comments

requested as to proposed no significant

hazards consideration: Yes (61 FR 27371

dated May 31, 1996). 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 July 1, 1996, but

indicated that if the Commission makes

a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendments. The May 31 and

June 5, 1996, letters provided additional

information that did not change the

scope of the May 28, 1996, application.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and a final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

June 17, 1996.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

October 24, 1994, as supplemented

August 31, 1995 and February 8, 1996.

The August 31, 1995 and February 8,

1996, letters provide clarification

information. The new information

changed the scope of the October 24,

1994, letter and was re-noticed on May

8, 1996, but did not change the initial

no significant hazards consideration

determination.

Brief description of amendment: The

proposed amendment would revise the

TS to allow the relocation of TS 3/

4.11.2.6, Gas Storage Tanks; and the

associated Bases in the TS to licenseecontrolled documents.

Date of issuance: June 12, 1996

Effective date: June 12, 1996

Amendment No.: 64

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60379). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated
34906 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

June 12, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

December 21, 1995

Brief description of amendments: The

amendments delete the requirement to

place the reactor mode switch in the

Shutdown position if a stuck open

safety/relief valve can not be closed

within 2 minutes. The operator will still

be required to scram the reactor if

suppression pool average water

temperature reaches 110 degrees

Fahrenheit or greater. The amendment

also includes editorial changes to the

index pages.

Date of issuance: June 18, 1996

Effective date: Immediately, to be

implemented within 60 days

Amendment Nos.: 113 and 98

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20844) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 18, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

July 18, 1994, as supplemented by letter

dated October 9, 1994

Brief description of amendments: The

amendments revise the current

combined Technical Specifications (TS)

for Units 1 and 2 by separating them

into individual volumes for Unit 1 and

Unit 2. In addition to the changes

required by the TS split, some

administrative and editorial changes

were made, such as the correction of

typographical errors and the deletion of

unnecessary blank pages.

Date of issuance: June 12, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 148 and 142

Facility Operating License Nos. NPF35 and NPF-52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 14, 1994 (59 FR

47166) The October 9, 1995 and June 6,

1996, letters provided clarifying

information that did not change the

scope of the July 18, 1994, application

and the initial proposed no significant

hazards consideration. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 12, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company, Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

February 22, 1996

Brief description of amendment: The

amendment increased the safety

function lift setpoint tolerances for the

safety and relief valves that are listed in

Surveillance Requirement 3.4.4.1 (Page

3.4-10) of the Technical Specifications

(TSs) for the Grand Gulf Nuclear

Station, Unit 1. The tolerances were

increased from the current plus or

minus 1 percent of the safety function

(i.e., safety relief valve) lift setpoint to

plus or minus 3 percent.

Date of issuance: June 12, 1996

Effective date: June 12, 1996

Amendment No: 123

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13524)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 12, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

May 1, 1996

Brief description of amendment: The

amendment revises the Operating

License and Technical Specifications

(TS) to implement 10 CFR Part 50,

Appendix J - Option B, by referring to

Regulatory Guide 1.163, ‘‘PerformanceBased Containment Leak-Test Program.’’

Specifically, changes have been made to

paragraph 2.D of the Operating License;

TS Section 1.1, ‘‘Definitions;’’ TS

3.6.1.1, ‘‘Primary Containment;’’ TS

3.6.1.1, ‘‘Primary Containment Air

Locks;’’ TS 3.6.1.3, ‘‘Primary

Containment Isolation Valves (PCIVs);’’

and TS Section 5.5, ‘‘Programs and

Manuals.’’

Date of issuance: June 21, 1996

Effective date: June 21, 1996

Amendment No.: 105

Facility Operating License No. NPF62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 21, 1996 (61 FR 25708)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 21, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

January 25, 1996

Brief description of amendment: The

amendment revises Technical

Specification 3/4.3.3, Emergency Core

Cooling System Actuation

Instrumentation, to more clearly define

when, during shutdown and refueling,

the Loss of Voltage and Degraded

Voltage relays for the Loss of Power

actuation trip functions are required to

be operable.

Date of issuance: June 10, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 72

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20851) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 10, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126
Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices 34907

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

January 5, 1996, as supplemented on

May 31, 1996

Brief description of amendment: The

amendment implements the guidance of

Generic Letter 93-08 by relocating

Tables 3.3-2, ‘‘Reactor Protective

Instrumentation Response Times’’ and

3.3-5, ‘‘Engineered Safety Features

Response Times’’ from the Technical

Specifications to the Millstone Unit No.

2 Technical Requirements Manual

(TRM). In accordance with Generic

Letter 93-08, the Limiting Conditions for

Operations for Technical Specifications

3.3.1.1, 3.3.2.1, and 3.7.1.6 are revised

to eliminate their references to the

aforementioned tables. The amendment

also revises Bases 3/4.3.1 and 3/4.3.2 to

reference that the instrument response

times are located in the TRM and that

these tables in the TRM are now

controlled under 10 CFR 50.59. The

amendment also removes a cyclespecific note from Tables 3.3-3 and 3.3-

4.

Date of issuance: June 10, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 198

Facility Operating License No. DPR65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5816) The May 31, 1996, letter provided

additional information that did not

change the scope of the January 5, 1996,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 10, 1996.

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, CT 06360 and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut

06385

PECO Energy Company, Public Service

Electric and Gas Company Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station, Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments:

February 15, 1996

Brief description of amendments: The

amendments change the Technical

Specifications to implement 10 CFR Part

50, Appendix J, Option B, by creating

Technical Specification Section 5.5.12,

‘‘Primary Containment Leakage Rate

Testing Program,’’ which refers to

Regulatory Guide 1.163, ‘‘PerformanceBased Containment Leakage-Test

Program.’’

Date of issuance: June 18, 1996

Effective date: Both units, as of date

of issuance, to be implemented by June

28, 1996.

Amendments Nos.: 214 and 219

Facility Operating License Nos. DPR44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13531)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 18, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

April 25, 1996

Brief description of amendments: The

amendments relocate Technical

Specification Traversing In-Core Probe

System Limiting Condition for

Operation 3/4.3.7.7 and its Bases 3/

4.3.7.7, to the Limerick Generating

Station Technical Requirements

Manual, and modify Note (f) of TS Table

4.3.1.1-1.

Date of issuance: June 11, 1996

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment Nos.: 117 and 79

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20840) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 11, 1996 No

significant hazards consideration

comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

March 14, 1996

Brief description of amendment: The

proposed changes would allow a onetime extension of the intervals for the

steam generator tube inspection that is

due in July 1996.

Date of issuance: June 19, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 166

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20854) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 19, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610

Tennessee Valley Authority, Docket

Nos. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of application for amendment:

February 28, as supplemented April 15,

and June 3, 1996.

Brief description of amendment: The

proposed amendment would revise the

Technical Specifications (TS) to

increase the surveillance intervals for

ice bed weight sampling and flow

passage inspection from 9 months to 18

months. The TS would also be changed

to provide an increased ice sublimation

allowance, associated with the

increased surveillance interval, by

increasing the minimum total ice weight

from 2,360,875 pounds to 2,403,800

pounds (1214 pounds/basket to 1236

pounds/basket).

Date of issuance: June 13, 1996

Effective date: June 13, 1996

Amendment No.: 2

Facility Operating License No. NPF90: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR 15998)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 13, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402
34908 Federal Register / Vol. 61, No. 129 / Wednesday, July 3, 1996 / Notices

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:

April 26, 1996

Brief description of amendment: The

amendment corrected minor technical

and administrative errors in the

Improved Technical Specifications prior

to its implementation.

Date of issuance: June 18, 1996

Effective date: June 18, 1996

Amendment No.: 85

Facility Operating License No. NPF58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 9, 1996 (61 FR 21213) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 18, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment requests: April 25

(TXX-94119) and August 12, 1994 (TXX94216), as supplemented by letters

dated February 15 (TXX-96055), March

7 (TXX-96078), and April 11, 1996

(TXX-96111).

Brief description of amendments:

These amendments modified the

Administrative Controls specifications,

relocate/remove requirements that are

adequately controlled by existing

regulations other than 10 CFR 50.36 and

the technical specifications. Guidance

on the proposed changes was developed

by NRC and provided in the Standard

Technical Specifications for

Westinghouse Plants, NUREG-1431. The

changes also update unit staff

qualification requirements to Regulatory

Guide 1.8, Revision 2.

Date of issuance: June 12, 1996

Effective date: June 12, 1996, to be

implemented witnin 60 days.

Amendment Nos.: Unit 1 -

Amendment No. 50; Unit 2 -

Amendment No. 36

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39599)

and September 28, 1994 (59 FR 49439).

The additional information contained in

the supplemental letters dated February

15, March 7, and April 11, 1996, were

clarifying in nature and thus, within the

scope of the initial notice and did not

affect the staff’s proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 12, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, Texas 76019

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: March

12, 1996 (TXX-96008)

Brief description of amendments: The

amendments revised the Technical

Specifications to reflect the approval for

the licensee to use of the new

Containment Leakage Rate Testing

Program as required by 10 CFR Part 50,

Appendix J, Option B for Comanche

Peak Steam Electric Station, Units 1 and

2. Implementation of the new

performance based leakage rate testing

program will be based on the guidance

provided by Regulatory Guide 1.163,

September 1995.

Date of issuance: June 13, 1996

Effective date: June 13, 1996, to be

implemented within 60 days

Amendment Nos.: 51 and 37

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR 15999)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 13, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

September 9, 1994, as superseded by

letter dated July 25, 1995, and

subsequently supplemented by letters

dated February 28, 1996, and April 9,

1996.

Brief description of amendment: The

amendment would revise TS 3/4.8.1 and

its associated Bases to improve the

overall emergency diesel generator

reliability and availability.

Date of issuance: June 17, 1996

Effective date: June 17, 1996, to be

implemented within 30 days of the date

of issuance.

Amendment No.: 112

Facility Operating License No. NPF30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45188)

The February 28, 1996, and April 9,

1996 supplemental letters 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

June 17, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251

Virginia Electric and Power Company,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia.

Date of application for amendments:

April 15, 1996

Brief description of amendments:

These amendments would revise the

Technical Specifications to indicate that

the quadrant power tilt ratio

requirements are applicable only at

power levels greater than 50% of rated

core power.

Date of issuance: June 7, 1996

Effective date: June 7, 1996

Amendment Nos.: 210 and 210

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20860) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 7, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185

Dated at Rockville, Maryland, this

26th day of June 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation

[Doc. 96–16879 Filed 7–2–96; 8:45 am]

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