Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40013

license with respect to such 12.58-

percent ownership interest and that the

transfer, subject to the conditions set

forth herein, is otherwise consistent

with applicable provisions of law,

regulations, and orders issued by the

Commission.

III

By August 30, 1996, any person

adversely affected by this order may file

a request for a hearing with respect to

issuance of the order. Any person

requesting a hearing shall set forth with

particularity how such person’s interest

is adversely affected by this order and

shall address the criteria set forth in 10

CFR 2.714(d).

If a hearing is to be held, the

Commission will issue an order

designating the time and place of such

hearing.

If a hearing is held concerning this

order, the issue to be considered at any

such hearing will be whether this order

should be sustained.

Any request for a hearing 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. Copies should also be sent to the

Office of the General Counsel and to the

Director, Office of Nuclear Reactor

Regulation, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and to Gerald Charnoff, Esquire,

of Shaw, Pittman, Potts and Trowbridge,

2300 N Street NW., Washington, DC

20037.

IV

Accordingly, pursuant to Sections

161b, 161i, and 184 of the Atomic

Energy Act of 1954, as amended, 42

U.S.C. §§ 2201(b), 2201(i), and 2234;

and 10 CFR 50.80, It is hereby ordered

that the Commission consents to the

proposed transfer of the license

described herein from Ohio Edison to

OES, subject to the following: Should

the transfer not be completed by

September 30, 1996, this order will

become null and void, unless upon

application and for good cause shown

this date is extended.

This order is effective upon issuance.

For further details with respect to this

action, see the application for transfer

dated December 28, 1995, under cover

of letter dated December 29, 1995,

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 located at

the Perry Public Library, 3753 Main

Street, Perry, Ohio.

For the Nuclear Regulatory Commission.

Dated at Rockville, Maryland this 25th day

of July 1996.

William T. Russell,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 96–19436 Filed 7–30–96; 8:45 am]

BILLING CODE 7590–01–P

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 July 6, 1996,

through July 19, 1996. The last biweekly

notice was published on July 17, 1996

(61 FR 37295).

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 30, 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
40014 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

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.

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan

Date of amendment request: March

25, 1996 (NRC-96-0003)

Description of amendment request:

The proposed amendment would

modify the charcoal testing standards

for the Control Room Emergency

Filtration System (CREFS) and the

Standby Gas Treatment System (SGTS)

to the current industry standard. The

changes affect Surveillance

Requirements (SRs) 4.6.5.3.b.2, 4.6.5.3.c,

4.7.2.1.c.2, and 4.7.2.1.d in Technical

Specifications (TS) 3/4.6.5.3 ‘‘Standby

Gas Treatment System’’ and TS 3/4.7.2

‘‘Control Room Emergency Filtration

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 TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. By providing an improved

protocol for charcoal testing the proposal

provides greater assurance that the installed

charcoal can perform its design function and,

thus, the consequences of evaluated

accidents remain valid. The method of

laboratory analysis has no effect upon how

the plant is operated, including the method

of sample removal. Therefore, the probability

[or consequences] of any evaluated accident

is unchanged.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. The proposal has no effect on the

manner of plant operation. The proposal does

not involve any change to the plant design.

Therefore, the change creates no new

accident modes.
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40015

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

By providing an improved protocol for

charcoal testing the proposal acts to maintain

existing safety margins. The change to the

SGTS charcoal acceptance criteria also acts to

ensure that the existing margins, as discussed

in Regulatory Guide 1.52, Revision 2 [Design,

Testing and Maintenance Criteria for PostAccident Engineered Safety-Feature

Atmosphere Cleanup System Air Filtration

and Adsorption Units of Light-Water-Cooled

Nuclear Power Plants], are maintained.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226

NRC Project Director: Mark Reinhart,

Acting

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: June 18,

1996

Description of amendment request:

For Beaver Valley Power Station, Unit

No. 1 (BVPS-1) only, the proposed

amendment would revise Technical

Specification (TS) 3.4.5 and associated

Bases; the Bases for TS 3.4.6.2 would

also be revised. The proposed changes

are editorial in nature and are intended

to provide consistency between the TSs

and associated Bases. Index page XIX

would be revised to reflect the revision

of page numbers for TS Tables 4.4-1 and

4.4-2 due to shifting of text.

For Beaver Valley Power Station, Unit

No. 2 (BVPS-2) only, the proposed

amendment would implement a voltagebased repair criteria for steam generator

tubes similar to the changes approved

for BVPS-1 by License Amendment No.

198. The proposed changes are intended

to reflect the guidance provided in NRC

Generic Letter 95-05, ‘‘Voltage-Based

Repair Criteria for Westinghouse Steam

Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.’’

The proposed changes would revise TSs

3.4.5 and 3.4.6.2 and associated Bases.

TS Table 4.4-2 would be revised to

reference TS 6.6 for reporting

requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Tube burst criteria are inherently satisfied

during normal operating conditions due to

the proximity of the tube support plate (TSP).

Test data indicates that tube burst cannot

occur within the TSP, even for tubes which

have 100% throughwall electric discharge

machining notches, 0.75 inch long, provided

that the TSP is adjacent to the notched area.

Since tube-to-TSP proximity precludes tube

burst during normal operating conditions,

use of the criteria must retain tube integrity

characteristics which maintain a margin of

safety of 1.43 times the bounding faulted

condition, main steamline break (MSLB)

pressure differential. The Regulatory Guide

(RG) 1.121 criterion requiring maintenance of

a safety factor of 1.43 times the MSLB

pressure differential on tube burst is satisfied

by 7/8’’ diameter tubing with bobbin coil

indications with signal amplitudes less than

8.6 volts, regardless of the indicated depth

measurement.

The upper voltage repair limit (VURL) will

be determined prior to each outage using the

most recently approved NRC database to

determine the tube structural limit (VSL). The

structural limit is reduced by allowances for

nondestructive examination (NDE)

uncertainty (VNDE) and growth (VGR) to

establish VURL. Using the Generic Letter (GL)

95-05 NDE and growth allowances for an

example, the NDE uncertainty component of

20% and a voltage growth allowance of 30%

per full power year can be utilized to

establish a VURL of 5.7 volts. The 20% NDE

uncertainty represents a square-root-sum-ofthe-squares (SRSS) combination of probe

wear uncertainty and analyst variability. The

degradation growth allowance should be an

average growth rate or 30% per effective full

power year, whichever is larger.

Relative to the expected leakage during

accident condition loadings, it has been

previously established that a postulated

MSLB outside of containment but upstream

of the main steam isolation valve (MSIV)

represents the most limiting radiological

condition relative to the plugging criteria. In

support of implementation of the revised

plugging limit, analyses will be performed to

determine whether the distribution of

cracking indications at the tube support plate

intersections during future cycles are

projected to be such that primary-tosecondary leakage would result in postulated

site boundary and control room doses

exceeding 10 CFR 100, 10 CFR 50 Appendix

A, and GDC-19 [General Design Criterion-19]

requirements, respectively. A separate

calculation has determined the maximum

allowable MSLB leakage limit in a faulted

loop. This limit was calculated using the

technical specification reactor coolant system

(RCS) Iodine-131 activity level of 1.0

microcuries per gram dose equivalent Iodine131 and the recommended Iodine-131

transient spiking values consistent with

NUREG-0800. The projected MSLB leakage

rate calculation methodology prescribed in

Section 2.b of GL 95-05 will be used to

calculate the end-of-cycle (EOC) leakage.

Projected EOC voltage distribution will be

developed using the most recent EOC eddy

current results and considering an

appropriate voltage measurement

uncertainty. The log-logistic probability of

leakage correlation will be used to establish

the MSLB leakrate used for comparison with

the faulted loop allowable limit. Therefore, as

implementation of the voltage-based repair

criteria does not adversely affect steam

generator tube integrity and implementation

will be shown to result in acceptable dose

consequences, the proposed amendment does

not result in any increase in the probability

or consequences of an accident previously

evaluated in the Updated Final Safety

Analysis Report (UFSAR).

The proposed changes to the BVPS-1

Index, Specifications and associated Bases

and the proposed change to BVPS-2 Table

4.4-2 are editorial in nature. Therefore, these

changes do not involve an increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

Implementation of the proposed steam

generator tube voltage-based repair criteria

does not introduce any significant changes to

the plant design basis. Use of the voltagebased repair criteria does not provide a

mechanism which could result in an accident

outside of the region of the tube support plate

elevations as no outside diameter stress

corrosion cracking (ODSCC) is occurring

outside the thickness of the tube support

plates. Neither a single or multiple tube

rupture event would be expected in a steam

generator in which the plugging limit has

been applied (during all plant conditions).

Duquesne Light Company will implement

a maximum primary-to-secondary leakage

rate limit of 150 gpd [gallons per day] per

steam generator to help preclude the

potential for excessive leakage during all

plant conditions. The RG 1.121 criterion for

establishing operational leakage rate limits

that require plant shutdown are based upon

leak-before-break considerations to detect a

free span crack before potential tube rupture

during faulted plant conditions. The 150 gpd

limit provides for leakage detection and plant

shutdown in the event of the occurrence of

an unexpected single crack resulting in

leakage that is associated with the longest

permissible crack length. RG 1.121

acceptance criteria for establishing operating

leakage limits are based on leak-before-break

considerations such that plant shutdown is

initiated if the leakage associated with the

longest permissible crack is exceeded.

The single through-wall crack lengths that

result in tube burst at 1.43 times the MSLB

pressure differential and the MSLB pressure

differential alone are approximately 0.57

inch and approximately 0.84 inch,

respectively. A leak rate of 150 gpd will

provide for detection of approximately 0.41

inch long cracks at nominal leak rates and

approximately 0.62 inch long cracks at the

lower 95% confidence level leak rates. Since

tube burst is precluded during normal
40016 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

operation due to the proximity of the TSP to

the tube and the potential exists for the

crevice to become uncovered during MSLB

conditions, the leakage from the maximum

permissible crack must preclude tube burst at

MSLB conditions. Thus, the 150 gpd limit

provides for plant shutdown prior to

reaching critical crack lengths for MSLB

conditions using the lower 95% leakrate

data. Additionally, this leak-before-break

evaluation assumes that the entire crevice

area is uncovered during blowdown. Partial

uncovery will provide benefit to the burst

capacity of the intersection. Analyses have

shown that only a small percentage of the

TSPs are deflected greater than the TSP

thickness during a postulated MSLB.

As steam generator tube integrity upon

implementation of the voltage-based repair

criteria continues to be maintained through

inservice inspection and primary-tosecondary leakage monitoring, the possibility

of a new or different kind of accident from

any accident previously evaluated is not

created.

The proposed change to BVPS-1 Index,

Specifications and associated Bases and the

proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. These changes do not

change the performance of plant systems,

plant configuration or method of operating

the plant.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The use of the voltage-based repair criteria

at BVPS-2 maintains steam generator tube

integrity commensurate with the criteria of

RG 1.121. This guide describes a method

acceptable to the Commission for meeting

GDCs 14, 15, 30, 31, and 32 by reducing the

probability or the consequences of steam

generator tube rupture. This is accomplished

by determining the limiting conditions of

degradation of steam generator tubing, as

established by inservice inspection, for

which tubes with unacceptable cracking

should be repaired or removed from service.

Upon implementation of the proposed

criteria, even under the worst case

conditions, the occurrence of ODSCC at the

tube support plate elevations is not expected

to lead to a steam generator tube rupture

event during normal or faulted plant

conditions. The EOC distribution of crack

indications at the tube support plate

elevations will be confirmed to result in

acceptable primary-to-secondary leakage

during all plant conditions and that

radiological consequences remain within the

licensing basis.

In addressing the combined effects of lossof-coolant-accident (LOCA) + safe shutdown

earthquake (SSE) on the steam generator

component (as required by GDC 2), it has

been determined that tube collapse may

occur in the steam generators at some plants.

This is the case as the tube support plates

may become deformed as a result of lateral

loads at the wedge supports at the periphery

of the plate due to the combined effects of

the LOCA rarefaction wave and SSE loadings.

Then, the resulting pressure differential on

the deformed tubes may cause some of the

tubes to collapse. There are two issues

associated with steam generator tube

collapse. First, the collapse of steam

generator tubing reduces the RCS flow area

through the tubes. The reduction in flow area

increases the resistance to flow of steam from

the core during a LOCA which, in turn, may

potentially increase peak clad temperature.

Second, there is a potential that partial

through-wall cracks in tubes could progress

to complete through-wall cracks during tube

deformation or collapse.

The results of an analysis using the larger

break inputs show that the LOCA loads were

found to be of insufficient magnitude to

result in steam generator tube collapse or

significant deformation. Since the leakbefore-break methodology is applicable to the

reactor coolant loop piping, the probability of

breaks in the primary loop piping is

sufficiently low that they need not be

considered in the structural design of the

plant. The limiting LOCA event becomes the

pressurizer spray line break. Analysis results

have demonstrated that no tubes were subject

to deformation or collapse. No tubes have

been excluded from application of the subject

voltage-based steam generator tube repair

criteria.

Addressing RG 1.83 considerations,

implementation of the voltage-based repair

criteria is supplemented by: enhanced eddy

current inspection guidelines to provide

consistency in voltage normalization, the

bobbin coil inspection will include 100% of

the hot-leg TSP intersections and cold-leg

intersections down to the lowest cold-leg

TSP with known ODSCC, the determination

of the TSPs having ODSCC will be based on

the performance of at least 20% random

sampling of tubes inspected over their full

length, and rotating pancake coil inspection

requirements for the larger indications left

inservice to characterize the principal

degradation as ODSCC.

As noted previously, implementation of

the tube support plate intersection voltagebased repair criteria will decrease the

number of tubes which must be repaired. The

installation of steam generator tube plugs

reduces the RCS flow margin. Thus,

implementation of the voltage-based repair

criteria will maintain the margin of flow that

would otherwise be reduced in the event of

increased tube plugging.

The proposed change to the BVPS-1 Index,

Specifications and associated Bases and the

proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. These changes will not

reduce the margin of safety because they

have no impact on any safety analysis

assumptions.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a significant reduction in margin

with respect to plant safety as defined in the

UFSAR or any BASES of the plant technical

specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: May 9,

1996

Description of amendment request:

The proposed amendment changes both

technical and administrative

requirements associated with station

batteries. The proposed changes are

modeled after ‘‘Standard Technical

Specifications - Babcock and Wilcox

Plants,’’ NUREG-1430 and Nuclear

Energy Institute guidance, ‘‘IEEE

Recommended Practice for

Maintenance, Testing, and Replacement

of Vented Lead-Acid Batteries for

Stationary Applications,’’ IEEE Std 450-

1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The switchyard 125V DC control power

source requirements do not meet the criteria

for inclusion in Technical Specifications

(TSs) as evaluated with respect to the

selection criteria of 10 CFR 50-36. These

control power sources are not assumed to

mitigate accident or transient events. The

effects of a loss of these control power

sources are enveloped by the Loss of Offsite

Power (LOOP) event and relocation is

considered to have a non-significant impact

on the probability or severity of a LOOP

event. These requirements will be relocated

from the TSs to an appropriate

administratively controlled document and

maintained pursuant to 10 CFR 50.59.

Proposed changes incorporating the

requirements of TS 3.7.1.D, 3.7.2.E, 3.7.2.F,

and 3.7.2.A, as related to the DC electrical

power subsystems in the new TS 3.7.3 results

in a more stringent requirement for the ANO1 TSs in that reductions to lower conditions

of operation in shorter periods of time are

now required. These more stringent

requirements are not assumed to be initiators

of any analyzed events and will not alter

assumptions relative to mitigation of accident

or transient events.

Proposed changes incorporating TS 3.7.4.

requirements for the station batteries

allowing the battery parameters to be outside
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40017

the limits of the Battery Inspection Program

for 31 days do not result in an increase in the

frequency of consequences of any analyzed

accident, as the actions require more frequent

checks of other parameters to ensure battery

capability during this 31 day period. The

Battery Inspection Program also requires

evaluations to determine battery operability

in the event these limits are exceeded. If an

evaluation shows the battery is incapable of

performing its design basis function, that DC

electrical subsystem will be declared

inoperable, and the appropriate actions

taken.

Proposed changes to allow the use of float

current in lieu of specific gravity incorporate

current industry guidance on operability

measures for station batteries, as stated in

IEEE-450, ‘‘IEEE Recommended Practice for

Maintenance, Testing, and Replacement of

Vented Lead-Acid Batteries for Stationary

Applications.’’ This Surveillance

Requirement is not considered to initiate or

mitigate any analyzed accident.

The proposed incorporation of a Battery

Inspection Program relocates maintenance

requirements from the TSs to a program

under 10 CFR 50.59 control and allows the

TSs to concentrate on those items required to

ensure battery operability. These relocated

requirements are not considered to be

initiators of any analyzed accident. Battery

operability is assured by the combination of

TS Surveillance Requirements and Battery

Inspection Program maintenance

requirements based on IEEE-450 guidance.

Proposed changes in Surveillance

Requirements and Frequencies reflect current

industry guidance on maintenance and

testing of the station batteries. These

requirements, in themselves, are not

considered to be initiators of any analyzed

accident condition. Although some

frequencies have been extended, continued

performance of maintenance activities in

accordance with IEEE-450, in addition to the

required Surveillance Requirements, ensures

that corrective maintenance can be

performed prior to a condition challenging an

operability limit.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed changes do not change the

design, configuration, or method of operation

of the plant.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

Relocation of the switchyard 125V DC

control power source requirements has no

impact on any safety analysis assumptions.

In addition, the requirements associated with

these control power sources are relocated to

an owner controlled document for which

future changes will be evaluated pursuant to

the requirements of 10 CFR 50.59.

Proposed changes incorporating the

requirements of TS 3.7.1.D, 3.7.2.E, 3.7.2.F,

and 3.7.2.A, as related to the DC electrical

power subsystems, in the new TS 3.7.3

impose more stringent requirements than

previously specified for ANO-1.

Proposed changes incorporating TS 3.7.4

requirements for the station batteries

allowing the battery parameters to be outside

the limits of the Battery Inspection Program

for 31 days may involve an incremental

reduction in the margin of safety since the

battery may be in a slightly degraded state.

However, this reduction is not considered

significant in that the associated actions

require more frequent checks of other

parameters to ensure battery capability

during this 31 day period. The attery

Inspection Program also requires evaluations

to determine battery operability in the event

these limits are exceeded.

If an evaluation shows the battery is

incapable of performing its design basis

function, that DC electrical subsystem will be

declared inoperable, and the appropriate

actions taken.

The proposed change to allow the use of

float current in lieu of specific gravity as a

measure of battery operability is expected to

result in a more representative measure of

operability. IEEE-450 states that specific

gravity may not be an appropriate measure of

battery capability following addition of

electrolyte or when the battery is on recharge

following a discharge.

Proposed incorporation of a Battery

Inspection Program relocates maintenance

requirements from the TSs to a program

under 10 CFR 50.59 controls and allows the

TSs to concentrate on those items required to

ensure battery operability. The relocation of

these requirements is not considered to be a

reduction in the margin of safety. Battery

operability is assured by the combination of

TS Surveillance Requirements and Battery

Inspection Program maintenance

requirements based on IEEE-450 guidance.

Proposed changes in Surveillance

Requirements and Frequencies reflect current

industry guidance on maintenance and

testing of the station batteries. Although

some frequencies have been extended,

continued performance of maintenance

activities in accordance with IEEE-450, in

addition to the required Surveillance

Requirements, ensures that corrective

maintenance can be performed prior to a

condition challenging an operability limit.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 27,

1996

Description of amendment request:

The proposed amendment will modify

Technical Specification 3/4.3.3.6,

‘‘Accident Monitoring Instrumentation,’’

based on the Combustion Engineering

improved Standard Technical

Specifications (STS) issued by the NRC

as NUREG 1432. The amendment will

also revise the Technical Specification

(TS) to include Accident Monitoring

Instrumentation as recommended by

Regulatory Guide (RG) 1.97, Revision 3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change deletes all non-Type

A and non-Category 1 instruments from the

requirements of TS 3/4.3.3.6, ‘‘Accident

Monitoring Instrumentation.’’ Type A

variables provide the primary information

required to permit the control room operators

to take specific manually controlled actions,

for which no automatic control is provided,

that are required for safety systems to

accomplish their safety functions during a

DBA [Design Basis Accident]. Category 1,

non-Type A variables are important in

reducing public risk and are retained in TS

because they are intended to assist operators

in minimizing the consequences of accidents.

Category 2 instruments are generally

designated for indicating system operating

status and are not designated as essential key

variables necessary for the safe shutdown of

the plant. The proposed change preserves the

safety requirements of RG 1.97, Revision 3,

and will not adversely affect any material

condition of the plant that could directly

contribute to causing or mitigating the affects

of an accident.

The proposed change also adds two

parameters to TS 3/4.3.3.6 which were

previously controlled administratively or per

another TS. Containment Pressure (Wide

Wide Range) is being added because it is a

Category 1 parameter required in addition to

Containment Pressure (Wide Range), which

is currently in the TS. Neutron Flux is being

added to distinguish the RG 1.97 channels

from the non-RG 1.97 channels and to

provide action and surveillance requirements

consistent with the other accident monitoring

instrumentation. These additions to TS 3/

4.3.3.6 contribute to the overall safety of the

plant and therefore in no way increase the

probability or consequences of an accident

previously evaluated.

Additionally, the proposed change also

extends the AOTs [Allowed Outage Times]

for TS 3/4.3.3.6 and replaces the HOT

SHUTDOWN requirement for the number of

OPERABLE channels being less than the

Required Number of channels with a Special

Report requirement. These changes are based
40018 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

on the relatively low probability of an

accident occurring which would require

these instruments, the passive nature of these

instruments, and alternate means of

monitoring available. This is consistent with

the CE improved STS and associated safety

analyses which have been approved and

issued by the NRC as NUREG 1432.

The remainder of the proposed change

provides enhancements and clarifications to

TS 3/4.3.3.6 which have no potential to

impact plant operations. No previous

accident scenario is changed, and initiating

conditions and assumptions remain as

previously analyzed. Therefore, the proposed

change will not involve a significant increase

in the probability or consequences of any

accident previously evaluated.

The proposed change will not alter the

operation of the plant or the manner in

which the plant is operated. No new or

different failure modes have been introduced.

TS 3/4.3.3.6 ensures the OPERABILITY of

essential Post Accident Monitoring

Instrumentation. This instrumen-tation

provides information to the control room

operators during an accident so that

appropriate actions can be taken to mitigate

the consequences of the accident. These

instruments are passive in nature in that no

critical automatic action is assumed to occur

from these instruments. Therefore, the

proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change revises TS 3/4.3.3.6

based on the information provided in CE

improved STS, NUREG 1432. The deletion

and addition of specific components from the

TS per this change is commensurate with the

safety significance of their associated

parameters. The proposed change ensures the

operability of the post accident monitoring

instrumentation which has been designated,

by RG 1.97 and Waterford 3’s associated

analysis, as essential for availability during

and following a DBA. The proposed change

preserves the single failure criteria required

for this instrumentation and maintains the

level of safety currently established in the

Technical Specifications. The proposed

change will not affect any physical protective

boundary. Therefore, the proposed change

will not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

N.W., Washington, D.C. 20005-3502

NRC Project Director: William D.

Beckner

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: July 17,

1996 (TSCR 242, Rev. 2)

Description of amendment request:

The proposed change to the Technical

Specifications would allow the

implementation of 10 CFR Part 50,

Appendix J, Option B. This application

supersedes the previously submitted

application dated February 23, 1996,

which was noticed in the Federal

Register on March 27, 1996 (61 FR

13526).

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:

GPU Nuclear has determined that this

TSCR involves no significant hazards

considerations as defined by NRC in 10 CFR

50.92.

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or occurrence or the

consequences of an accident of malfunction

of equipment important to safety as

previously evaluated in the Safety Analysis

Report.

The proposed change implements Option B

of 10 CFR 50, Appendix J on performance

based containment leakage testing. The

proposed change does not involve a change

to the plant design or operation. Therefore,

the proposed change does not affect any of

the parameters or conditions that contribute

to initiation of any of the analyzed accidents

or malfunctions. The proposed change does

not request an allowable extension of

containment testing. Therefore, a

hypothetical leak could remain undetected

for a greater period of time. This slight

increase in risk has been determined to be

insignificant as:

Type A Testing

NUREG 1493 [Performance-Based

Containment Leak Test Program] determined

that the effect of containment leakage on

overall accident risk is small as risk is

dominated by accident sequences that result

in the failure or bypass of the containment.

Industry wide PCILRTs [primary

containment integrated leak rate tests] have

demonstrated that only a small fraction of the

leaks discovered during testing exceeded

acceptance criteria, and that the leak rate has

been only marginally above the acceptable

limit. Only 3% of all leaks can be detected

only by PCILRT, therefore, only 3% of the

theoretical leaks are affected by the extension

to the Type A test interval. Experience at

Oyster Creek agrees with the industry wide

data in that the majority of the detected

leakage from the primary containment is

found through Type B and C testing. NUREG

1493 found that these observations, together

with the insensitivity of reactor accident risk

to the containment leakage rate, demonstrates

that increasing the Type A leakage test

intervals would have a minimal impact on

public risk.

Type B and C Testing

Penetrations are designed to ensure

reliability of the containment isolation

function. Type B penetrations use a double

passive seal (e.g. o-ring, gasket) and Type C

penetrations use a double isolation valve

design to ensure reliability of the isolation

function. Because valves perform the

isolation function actively, they are more

likely to fail on demand (e.g. failure to

completely close on demand). To address

this failure mode, Type C valves are

subjected to increased design constraints and

testing to ensure both acceptable leak rates

and stroke times. The proposed change does

not alter the installation, operation, operating

environment, or testing method of these

valves. Therefore, the proposed change does

not introduce any new component failure

modes, nor does it affect the probability of

occurrence of any existing evaluated failure

mode.

The failure of any single penetration

barrier (isolation valve or passive seal) does

not cause penetration failure. Therefore, a

double failure would have to occur to cause

a failure of the penetration and affect

containment. Additionally, the proposed

change does not change the acceptance

criteria for acceptable leakage testing.

The proposed change does not alter plant

design or operation, nor does it alter the

allowable maximum leakage rate limit. Thus,

the proposed change does not affect the

probability of occurrence nor the

consequences of any evaluated accident or

malfunction of equipment important to

safety.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of any accident or

malfunction different from any accident or

malfunction 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

could contribute to initiation of any

accidents. This change only involves the

reduction in Type A, B, and C test

frequencies, and the Type A test pressure.

Type A Testing

The only changes proposed to the Type A

testing are to frequency and test pressure. As

the proposed test pressure is greater than the

existing test pressure, no new type of

accident or malfunction is created, and the

increase in pressure provides an additional

margin of safety. The increase in surveillance

interval cannot introduce any new type of

accident or malfunction.

The PCILRT is presently performed at 20

psig. Performance of the PCILRT at

PGG5GA(35 psig) will provide a more direct

leak rate for analysis. Pa is the design

pressure of the torus (the drywell design

pressure is 44 psig, but the torus is non

isolable from the drywell). Therefore, Pa will

not create the possibility of the failure of the

torus due to overpressurization. No new

accident modes can be created by extending

the test intervals. No safety related functions
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40019

or components are altered as a result of this

change. Therefore, no new accident or

malfunction different from those evaluated in

the Safety Analysis Report can result due to

the increase in test pressure or increase in

surveillance interval.

Type B and C Testing

The proposed change only deals with the

frequency of performing Type B and C

testing. It does not change what components

are tested or the method of testing. There is

no proposed change to the design or

operation of the plant. Therefore, no new

accident or malfunction different from those

evaluated in the Safety Analysis Report can

result due to the increase in test pressure or

increase in surveillance interval.

3. Operation of the facility in accordance

with the proposed amendment would not

decrease the margin of safety as defined in

the bases of the Technical Specifications.

Type A Testing

Except for the method of defining the test

frequency and pressure at which the PCILRT

is performed, the methods for performing the

actual test are not changed. However, the

proposed change can increase the probability

that an increase in leakage could go

undetected for an extended period of time.

NUREG 1493 has determined that under

several different accident scenarios, the

increased risk of radioactivity release from

containment is negligible with the

implementation of these proposed changes.

Type B and C Testing

The proposed change only affects the

frequency of Type B and C testing. The

methods for performing the actual test are not

changed. The design or operation of Type B

and C components are not changed. The

proposed change will result in a longer

interval between tests of good performing

Type B and C components.

The margin of safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to containment leakage rate. The

containment isolation system is designed to

limit leakage to La, which is defined by the

Oyster Creek Technical Specifications to be

1.0 percent by weight of the containment air

at 35 psig per 24 hours. The limitation on

containment leakage rate is designed to

ensure the total leakage volume will not

exceed the value assumed in the accident

analyses at the peak accident pressure (Pa).

The margin of safety for the offsite dose

consequences of postulated accidents

directly related to the containment leakage

rate is maintained by meeting the 1.0 La

acceptance criteria. The La value is not being

modified by this proposed Technical

Specification change request.

Therefore, the margin of safety as defined

in the bases for the Technical Specification

will not be 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: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: June 28,

1996

Description of amendment request:

This amendment would allow

implementation of Option B to 10 CFR

Part 50, Appendix J, which permits

performance based determination of the

frequency of containment leak rate

testing.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

The proposed change has been evaluated

against the standards in 10 CFR 50.92 and

determined not to involve a significant

hazards consideration, in that the editorial

changes do not change the meaning or intent

of the technical specifications, and operation

of the facility in accordance with the

proposed amendment.

1. Would not involve a significant increase

in the probability of occurrence or the

consequences of an accident previously

evaluated, because the proposed changes are

either purely administrative changes

(involving format, wording, or reporting

requirements) or changes in containment

leakage test requirements (minor scope

changes or increased intervals between

containment leakage tests). None of these

changes are related to conditions which

cause accidents. The proposed changes do

not involve a change to the plant design or

operation.

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’

contributed to the technical bases for Option

B of 10 CFR 50 Appendix J. NUREG-1493

contains a detailed evaluation of the

expected leakage from containment and the

associated consequences. The increased risk

due to lengthening of the intervals between

leakage tests was also evaluated and found to

be acceptable. Using a statistical approach,

NUREG-1493 determined the increase in the

expected dose to the public from extending

the testing frequency to be extremely small.

2. Would not create the possibility of a new

or different kind of accident from any

accident previously evaluated, because the

testing or reporting requirements associated

with this change do not involve a physical

alteration of the plant design or changes in

the methods governing normal plant

operation. No safety related equipment or

safety related functions are altered as a result

of this change. As a result, the proposed

change does not affect any of the parameters

or conditions that could contribute to

initiation of any accidents.

3. Would not involve a significant

reduction in a margin of safety because the

proposed changes are either purely

administrative (involving format, wording, or

reporting requirements) or changes in

containment leakage test requirements

(minor scope changes or increased intervals

between containment leakage tests) such that

the allowable containment leakage rates

presently specified in the Technical

Specifications remain unchanged. The

Technical Specifications and the Reactor

Building Leakage Rate Testing Program will

ensure that containment system testing is

performed in full compliance with 10 CFR 50

Appendix J.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location:

Law/Government Publications Section,

State Library of Pennsylvania, (REGIONAL

DEPOSITORY) Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 1,

1995, as supplemented by letters dated

June 22, August 28, November 22, and

December 19, 1995, and January 4, 8

(two letters), and 23, June 27, and July

9, 1996.

Description of amendment request:

The proposed amendment would allow

extension of the standby diesel

generator allowed outage time to 14

days, and extension of the essential

cooling water loop and the essential

chilled water loop allowed outage times

to 7 days. The proposed change would

also add to Administrative Controls a

description of the Configuration Risk

Management Program (CRMP) used to

assess changes in core damage

probability resulting from applicable

plant configurations. This application

was previously published in the Federal
40020 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

Register on February 8, 1996, (61 FR

4805).

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 Standby Diesel Generators are not

accident initiators, therefore the increase in

Allowed Outage Times for this system does

not increase the probability of an accident

previously evaluated. The three train design

of the South Texas Project ensures that even

during the seven days the Essential Cooling

Water loop or the Essential Chilled Water

loop is inoperable there are still two

complete trains available to mitigate the

consequences of any accident. If the Essential

Cooling Water and the Essential Chilled

Water loops are operable during the 14 days

the Standby Diesel Generator is inoperable,

the Engineered Safety Features bus and

equipment in the train associated with the

inoperable Standby Diesel Generator will be

operable. This ensures that all three

redundant safety trains of the South Texas

Project design are operable. In addition the

Emergency Transformer will be available to

supply the Engineered Safety Features bus

normally supplied by the inoperable Standby

Diesel Generator. These actions will ensure

that the changes do not involve a significant

increase in the consequences of previously

evaluated accidents.

The addition of the Configuration Risk

Management Program to the Administrative

Section of the Technical Specifications does

not affect current accident analyses.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes affect only the

magnitude of the Standby Diesel Generator,

Essential Cooling Water and the Essential

Chilled Water Allowed Outage Times as

identified by the marked-up Technical

Specification. As indicated above, the

proposed change does not involve the

alteration of any equipment nor does it allow

modes of operation beyond those currently

allowed. Therefore, implementation of these

proposed changes does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes result in no

significant increase in core damage or large

early release frequencies. Three sets of PSA

[probabilistic safety assessment] results have

been presented to the NRC for the South

Texas Project. One submitted in 1989 from

the initial Level 1 PSA of internal and

external events with a mean annual average

CDF [core damage frequency] estimate of

1.7E-4, a second one submitted in 1992 to

meet the IPE [individual plant examination]

requirements from the Level 2 PSA/IPE with

a CDF estimate of 4.4E-5, and an update of

the PSA that was reported in the August 1993

Technical Specifications submittal with a

variety of CDF estimates for different

assumptions regarding the rolling

maintenance profile and different

combinations of modified Technical

Specifications. The South Texas Project PSA

was updated in March of 1995 to include the

NRC approved Risk-Based AOTs [allowed

outage times] and STIs [surveillance test

intervals], Plant Specific Data and

incorporate the Emergency Transformer into

the model. This update resulted in a CDF

estimate of 2.07E-5 per reactor year. When

the requested changes are modeled, the

resulting CDF estimate is 2.18E 10-5 (sic)

[2.18E-5] per reactor year. This corresponds

to 5.2% decrease in the Core Damage

Frequency calculated for the previously

submitted 21 Day AOT. The Large, Early

Release Frequency is quantified as 4.69E-07

per reactor year which represents a decrease

of 7.5% from the value calculated for the

previously submitted 21 Day AOT. Therefore,

it is concluded that there is no significant

reduction in the margin of safety.

Based on the above evaluation, the South

Texas Project has concluded that these

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 standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036-5869

NRC Project Director: William D.

Beckner

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: July 5,

1996

Description of amendment request:

The proposed Technical Specification

(TS) amendment would support

implementation of Noble Metal

Chemical Addition (NMCA) at the

Duane Arnold Energy Center (DAEC) as

a method to enhance the effectiveness of

Hydrogen Water Chemistry (HWC) in

mitigating Intergranular Stress

Corrosion Cracking (IGSCC) in Boiling

Water Reactor (BWR) vessel internal

components. The proposed amendment

would raise the reactor water

conductivity limit in STARTUP and

HOT SHUTDOWN only during the

application of NMCA. The reactor water

conductivity will be restored after the

NMCA.

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 amendment will not

significantly increase the probability or

consequences of any previously evaluated

accidents.

It is expected that during the NMCA

application period, the reactor water

conductivity will increase and exceed the

conductivity limit of 2.0 [micro]mhos/cm

specified in our current TS. Our current TS

requires that whenever the reactor is in

STARTUP or HOT SHUTDOWN Mode, the

conductivity shall not exceed 2.0

[micro]mhos/cm for more than 48 continuous

hours or be in HOT SHUTDOWN within the

next 12 hours and in COLD SHUTDOWN

within the following 24 hours.

The expected increase in conductivity is

due to the presence of noble metal chemistry

in the reactor water and is appropriate during

the [NMCA] application period. The

deposited layer of noble metals is beneficial

for mitigating IGSCC in reactor vessel

internal components. Other reactor water

chemistry parameters such as chloride and

sulfate are not expected to change; pH is

expected to change but not out of the

acceptable range. The reactor water

chemistry parameters will be analyzed to

ensure they are within the normal range, on

a frequency consistent with the existing TS,

Sections 4.6.B.2.c and 4.6.B.2.d when

conductivity is elevated during the NMCA

application.

During and after the application, the

Reactor Water Cleanup (RWCU) system will

continue to operate to remove the excess ions

from the reactor water and restore the reactor

water conductivity to the limit specified in

Section 3.6.B. Therefore, this proposed TS

amendment will not significantly increase

the probability or consequences of any

previously evaluated accidents.

2. The proposed TS amendment will not

create the possibility of a new or different

kind of accident. The proposed TS

amendment will only permit a higher value

of the reactor water conductivity limit during

the application period of NMCA. The

application is anticipated to increase the

reactor water conductivity.

During and after the application, the

RWCU system will continue to operate to

remove the excess ions and restore the

reactor water conductivity to the limit

specified in Section 3.6.B. As is discussed

above, the deposited layer of noble metals is

beneficial for mitigating IGSCC in reactor

vessel internal components. Therefore, this

proposed TS amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed TS amendment will only

permit a higher value of the reactor water

conductivity limit during the application

period of NMCA. The increase in
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40021

conductivity is anticipated during the

application and is appropriate. The deposited

layer of noble metals is beneficial for

mitigating IGSCC in reactor vessel internal

components. During and after the

application, the RWCU system will continue

to operate to remove the excess ions and

restore the reactor water conductivity to the

limit specified in Section 3.6.B. Therefore, no

margin of safety is reduced as a result of the

anticipated increase in conductivity due to

the addition of the known noble metals.

The NRC staff has reviewed the

licensee’s analysis and, based on

thisreview, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids,

Iowa 52401

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis, &

Bockius, 1800 M Street, NW.,

Washington, DC 20036-5869

NRC Project Director: Gail H. Marcus

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of amendment request: June 21,

1996

Description of amendment request:

The proposed amendment would

modify Section 5.7, ‘‘High Radiation

Areas,’’ of the ‘‘Administrative

Controls’’ section of the Clinton Power

Station technical specifications (TS).

The proposed changes include: (1)

allowing utilization of a Radiation Work

Permit (RWP) ‘‘or equivalent’’ to control

entry into a high radiation area; (2)

clarifying the example given in the TS

of individuals who are qualified in

radiation protection procedures; (3)

clarifying the requirements for when

specified access controls and barriers for

high radiation areas within large areas

like the containment must be

established; (4) clarifying that it is

acceptable for an RWP to specify a

maximum dose, i.e., a specified setpoint

on an alarming dosimeter in lieu of a

stay time for entry into a high radiation

area (where an individual could receive

a deep dose equivalent of 3000 mrem in

one hour); (5) eliminating the upper

dose limit for specifying the

applicability of the requirements of

Specification 5.7.1; (6) providing

additional flexibility regarding who may

control the keys to locked doors for

preventing unauthorized entry into high

radiation areas; (7) reorganizing TS

Sections 5.7.1, 5.7.2, and 5.7.3 into four

sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4);

and (8) making minor edits to enhance

readability.

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) None of the proposed changes involve

a significant increase in the probability or

consequences of any accident previously

evaluated.

The proposed changes do not change the

design or the operation of the plant. The

proposed changes are only related to the

control of access to high radiation areas for

the purpose of controlling dose to plant

personnel. Because no change to plant design

is proposed, there is no impact to any

accident mitigating system. Likewise,

because there is no proposed change to plant

operating procedures, plant operation is not

impacted. This proposed change does not

impact any accident scenario or the

previously calculated post-accident doses.

Therefore, the limits of 10 CFR 100 will

continue to be met. No probability or

consequence of any accident previously

evaluated is impacted by the proposed

changes to TS.

(2) None of the proposed changes create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed amendment is

administrative in nature and does not impact

directly or indirectly the design or the

operation of the Clinton Power Station, thus

no new accident can be created.

(3) None of the proposed changes involve

a significant reduction in a margin of safety.

There is no reduction to the margin of

safety because the operating limits and

functional capabilities of plant safety systems

are unaffected by the proposed changes to

administrative requirements. As noted

previously, the proposed changes do not

impact any accident analyses, including the

associated dose calculations. With respect to

controls for controlling operational dose to

plant personnel, the proposed changes are

intended to provide clarity and/or flexibility

with respect to the administration and

programmatic controls for controlling such

dose, and yet maintain an adequate margin

of safety for minimizing dose to site

personnel consistent with the requirements

of 10 CFR 20 and guidance of Regulatory

Guide 8.38.

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: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Attorney for licensee: Leah Manning

Stetzner, Vice President, General

Counsel, and Corporate Secretary, 500

South 27th Street, Decatur, Illinois

62525

NRC Project Director: Gail H. Marcus

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of amendment request: June 28,

1996

Description of amendment request:

The proposed amendment would allow

removal of the Inclined Fuel Transfer

System (IFTS) primary containment

blind flange while primary containment

is required to be operable. This will

provide flexibility to operate the IFTS

for the purpose of testing and exercising

the system during such conditions.

Primary containment integrity will be

provided by an alternate means while

the blind flange is removed. The change

would be incorporated via a provisional

note into Technical Specification (TS)

Surveillance Requirement 3.6.1.3.3,

associated with TS 3.6.1.3, ‘‘Primary

Containment Isolation Valves (PCIVs).’’

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 allows operation

of the IFTS while primary containment

operability is required. The proposed change

does not involve any modifications to plant

systems or design parameters or conditions

that contribute to the initiation of any

accidents previously evaluated. Therefore,

the proposed change cannot increase the

probability of any accident previously

evaluated.

The proposed change potentially affects

the leak-tight integrity of the containment

structure which is designed to mitigate the

consequences of a loss-of-coolant accident

(LOCA). The function of the primary

containment is to maintain functional

integrity during and following the peak

transient pressures and temperatures that

result from any LOCA. The primary

containment is designed to limit fission

product leakage following the design basis

LOCA. Because the proposed change does

not alter the plant design, only the extent of

the boundaries that provide primary

containment isolation for the IFTS

penetration, the proposed change does not

result in an increase in primary containment

leakage. However, temporarily using the IFTS

transfer tube and its attached appurtenances

as part of the primary containment boundary

(which have not been fabricated or installed

to exactly the same requirements as a fully

certified primary containment penetration)

can increase the probability that a LOCA

would challenge the pressure retaining

integrity of these components. Since the

subject components have been built to

withstand pressure, temperature, and seismic

conditions similar to those of the existing

penetration, they are judged to be an
40022 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

acceptable barrier to prevent the

uncontrolled release of post-accident fission

products for the purposes of this amendment

request.

Further, it has been shown that the largest

potential leakage pathway, the IFTS transfer

tube itself, would remain sealed by the depth

of water required to be maintained in the fuel

building fuel transfer pool. The transfer tube

drain line constitutes the other possible

leakage pathway, and will be required to be

capable of being isolated via administrative

control of the manual isolation valve in the

drain line. Additionally, due to the physical

relationships of the buildings and

components involved, any leakage from

either of these pathways is fully contained

within the boundaries of the secondary

containment and would be filtered by the

Standby Gas Treatment System prior to

release to the environment.

Based on the above, Illinois Power has

concluded that the proposed change will not

result in a significant increase in the

probability or consequences of any accident

previously evaluated.

(2) The proposed change does not involve

a change to the plant design or operation

(except when the IFTS is operated). As a

result, the proposed change does not affect

any of the parameters or conditions that

could contribute to the initiation of any

accidents. No new accident modes are

created by this change. Extending the

primary containment boundary to include

portions of the IFTS 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.

Based on the above, Illinois Power has

concluded that the proposed change will not

create the possibility of a new or different

kind of accident not previously evaluated.

(3) The request does not involve a

significant reduction in a margin of safety.

The proposed change only affects the extent

of a portion of the primary containment

boundary. Precautions will be taken to

administratively control the IFTS transfer

tube drain path so that the proposed change

will not increase the probability that an

increase in leakage from the primary

containment to the secondary containment

could occur.

The margin of safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to containment leakage rate. The

containment isolation system is designed to

limit leakage to La, which is defined by the

Clinton Power Station Technical

Specifications to be 0.65% of primary

containment air weight per day at the

calculated peak constant pressure (Pa). The

limitation on containment leakage rate is

designed to ensure that total leakage volume

will not exceed the value assumed in the

accident analyses at the peak accident

pressure (Pa). The margin of safety for the

offsite dose consequences of postulated

accidents directly related to the containment

leakage rate is maintained by meeting the 1.0

La acceptance criteria. The La value is not

being modified by this proposed technical

specification change. The IFTS will continue

to provide an acceptable barrier to prevent

containment leakage during a LOCA, and

therefore this change will not create a

situation causing the containment leakage

rate acceptance criteria to be violated.

As a result, Illinois Power has concluded

that the proposed change will 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: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Attorney for licensee: Leah Manning

Stetzner, Vice President, General

Counsel, and Corporate Secretary, 500

South 27th Street, Decatur, Illinois

62525

NRC Project Director: Gail H. Marcus

Indiana Michigan Power Company,

Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of amendment requests: June 11,

1996 (AEP:NRC:80027)

Description of amendment requests:

The proposed amendments would

remove from the technical specifications

(TS) certain requirements for

administrative controls, related to

quality assurance requirements, in

accordance with the guidance of NRC

Administrative Letter 95-06,≥Relocation

of Technical Specifications

Administrative Controls Related to

Quality Assurance.’’

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:

We have evaluated the proposed T/S

changes and have determined that the

changes should involve no significant

hazards consideration based on the criteria

established in 10 CFR 50.92(c). Operation of

Cook Nuclear Plant in accordance with the

proposed amendment will not satisfy any of

the following criteria:

(a) Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change does not involve any

physical alteration of plant configurations,

changes to setpoints, or operating parameters.

This proposed amendment is to relocate the

T/S requirements for administrative controls

that are related to quality assurance to the

QAPD [Quality Assurance Program

Description]. This is in accordance with the

guidance provided in AL 95-06. Also, the

relocated requirements and future changes

are controlled by 10 CFR 50.54(a) which

requires prior NRC approval for changes that

reduce the commitments in the program

description previously accepted by the NRC.

Therefore, there will be no significant

increase in the probability or consequences

of an accident previously evaluated.

(b) Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change does not involve any

physical alteration of plant configurations,

changes to setponts, or operating parameters.

This proposed amendment is to relocate the

T/S requirements for administrative controls

that are related to quality assurance to the

QAPD. This is in accordance with the

guidance provided in AL 95-06. Also, the

relocated requirements and future changes

are controlled by 10 CFR 50.54(a) which

requires prior NRC approval for changes that

reduce the commitments in the program

description previously accepted by the NRC.

Therefore, this proposed change does not

create the possibility of a new of different

kind of accident from any previously

analyzed.

(c) Involve a significant reduction in a

margin of safety.

The proposed change does not involve any

physical alteration of plant configurations,

changes to setpoints, or operating parameters.

This proposed amendment is to relocate the

T/S requirements for administrative controls

that are related to quality assurance to the

QAPD. This is in accordance with the

guidance provided in AL 95-06. Also, the

relocated requirements and future changes

are controlled by 10 CFR 50.54(a), which

requires prior NRC approval for changes that

reduce the commitments in the program

description previously accepted by the NRC.

Therefore, this proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Mark Reinhart,

Acting

Indiana Michigan Power Company,

Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien

County, Michigan

Date of amendment request: June 19,

1996 [AEP:NRC:1166AA]

Description of amendment request:

The proposed amendment would

modify the technical specifications (T/
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40023

S) to allow continued use of the 2-volt

steam generator (SG) tube plugging

criteria for future operating cycles as

discussed in NRC Generic Letter 95-05,

‘‘Voltage-Based Repair Criteria for the

Repair of Westinghouse Steam

Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.’’

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:

In accordance with the three factor test of

10 CFR 50.92(c), implementation of the

proposed license amendment is analyzed

using the following standards and found not

to: 1) involve a significant increase in the

probability or consequences of an accident

previously evaluated; 2) create the possibility

of a new or different kind of accident from

any accident previously evaluated; or 3)

involve a significant reduction in margin of

safety. Conformance of the proposed

amendment to the standards for a

determination of no significant hazards as

defined in 10 CFR 50.92 (three factor test) is

shown in the following paragraphs:

1) Operation of Cook Nuclear Plant Unit 1,

in accordance with the proposed license

amendment, does not involve a significant

increase in the probability or consequences

of an accident previously evaluated. Testing

of model boiler specimens for free span

tubing

(no TSP [tube support plate] restraint)

at room temperature conditions show

burst pressures in excess of 5000 psi for

indications of outer diameter stress

corrosion cracking [ODSCC] with

voltage measurements as high as 19

volts. Burst testing performed on pulled

tubes from Cook Nuclear Plant Unit 1

with up to a 2.02 volt indication shows

measured burst pressure in excess of

10,000 psi at room temperature. Burst

testing performed on pulled tubes from

other plants show burst pressures in

excess of 5,300 psi at room

temperatures. Correcting for the effects

of temperature on material properties

and minimum strength levels (as the

burst testing was done at room

temperature), tube burst resistance

significantly exceeds the safety factor

requirements of RG [Regulatory Guide]

1.121 [Bases for Plugging Degraded PWR

Steam Generatory Tubes]. As stated

earlier, tube burst criteria are inherently

satisfied during normal operating

conditions due to the proximity of the

TSP. Test data indicates that tube burst

cannot occur within the TSP, even for

tubes which have 100% throughwall

electric-discharge machined notches

0.75 inch long, provided the TSP is

adjacent to the notched area. Since tubeto-tube support plate proximity

precludes tube burst during normal

operating conditions, it follows that use

of the proposed plugging criteria must,

therefore, retain tube integrity

characteristics which maintain the RG

1.121 margin of safety of 1.43 times the

bounding faulted condition (steam line

break) pressure differential.

During a postulated main SLB [steamline

break], the TSP has the potential to deflect

during blowdown, thereby uncovering the

intersection. Based on the existing data base,

the RG 1.121 criterion requiring maintenance

of a safety factor of 1.43 times the SLB

pressure differential on tube burst is satisfied

by 7/8 inch diameter tubing with bobbin coil

indications with signal amplitudes less than

VSL, regardless of the indicated depth

measurement. A 2 volt plugging criteria

compares favorably with the current VSL (8.8

volt) structural limit, considering the

previously calculated growth rates for

ODSCC within Cook Nuclear Plant Unit 1

SGs. Considering a voltage growth

component of 0.8 volts (40% voltage growth

based on 2 volts BOC [beginning of cycle]

and a nondestructive examination

uncertainty of 0.40 volts (20% voltage

uncertainty based on 2 volts BOC), when

added to the BOC plugging criteria of 2 volts,

results in a bounding EOC [end of cycle]

voltage of approximately 3.2 volts for a cycle

operation. A 5.6 volt safety margin exists (8.8

- 3.2 volt EOC = 5.6 volt margin).

For the voltage/burst correlation, the EOC

structural limit is supported by a voltage of

8.8 volts. Using this VSL of 8.8 volts, a BOC

maximum allowable repair limit can be

established using the guidance of RG 1.121.

The BOC maximum allowable repair limit

should not permit a significant number of

EOC indications to exceed the VSL and

should assure that acceptable tube burst

probabilities are attained. By adding NDE

[nondestructive examination] uncertainty

allowances and an allowance for crack

growth to the repair limit, the structural limit

can be validated. The previous plugging

criteria submittal established the

conservative NDE uncertainty limit (VNDE) of

20% of the BOC repair limit. For consistency,

a 40% voltage growth allowance (VGR) to the

BOC repair limit is also included. This

allowance is extremely conservative for Cook

Nuclear Plant Unit 1. Therefore, the

maximum allowable upper voltage repair

limit VURL for BOC, based on the VSL of 8.8

volts, can be represented by the expression:

VURL + (VNDE x VURL) + (VGR x VURL) = 8.8

volts, or,

the maximum allowable BOC repair limit

can be expressed as,VURL = 8.8 volt structural

limit/1.6 = 5.5 volts.

This structural repair limit supports this

application for plugging criteria

implementation to repair bobbin indications

greater than 2 volts based on RPC [rotating

pancake coil] confirmation of the indication.

Conservatively, an upper limit of 5.5 volts

will be used to repair bobbin coil indications

which are above 2 volts but do not have

confirming RPC calls.

Relative to the expected leakage during

accident condition loadings, it has been

previously established that a postulated main

SLB outside of containment, but upstream of

the main steam isolation valve, represents the

most limiting radiological condition relative

to the plugging criteria. In support of

implementation of the plugging criteria, it

will be determined whether the distribution

of crack indications at the TSP intersections

at the EOC are projected to be such that

primary-to-secondary leakage would result in

site boundary doses within a small fraction

of the 10 CFR 100 guidelines. A separate

calculation has determined this allowable

SLB leakage limit to be 8.4 gpm. Although

not required by the Cook Nuclear Plant

design basis, this calculation uses the

recommended Iodine-131 transient spiking

values consistent with NUREG-0800

[Standard Review Plan], and the T/S reactor

coolant system activity limit of 1 micro curie

per gram dose equivalent Iodine-131. Control

room dose calculations were also performed

and found to be less limiting than the offsite

dose leakrate. Therefore, the more

conservative offsite dose leakrate is used. The

projected SLB leakage rate calculation

methodology prescribed in GL 95-05 and

WCAP 14277 [Steam Line Break Leak Rate

and Tube Burst Probability Analysis Methods

for Outside Diameter Stress Corrosion

Cracking at Tube Support Plate Intersections]

will be used to calculate EOC leakage, based

on actual EOC distributions and EOC

projected distributions. Due to the relatively

low voltage growth rates at Cook Nuclear

Plant Unit 1 and the relatively small number

of indications affected by the plugging

criteria, SLB leakage prediction per GL 95-05

is expected to be significantly less than the

permissible level of 8.4 gpm in the faulted

loop.

The inclusion of all intersections in the

leakage model, along with application of a

probability of detection of 0.6, will result in

extremely conservative leakage estimations.

Close examination of the available data

shows that indications of less than 2.8 volts

will not be expected to leak during SLB

conditions.

The proposed amendment does not result

in any increase in the probability or

consequences of an accident previously

evaluated within the cook Nuclear Plant Unit

1 Final Safety Analysis Report (FSAR).

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 SG tube

plugging criteria does not introduce any

significant changes to the plant design basis.

Use of the criteria does not provide a

mechanism which could result in an accident

outside of the region of the TSP elevations.

Neither a single nor a multiple tube rupture

event would, under any plant conditions, be

expected in a SG in which the plugging

criteria has been applied. Specifically, we

will continue to implement a maximum

leakage rate limit of 150 gpd (0.1 gpm) per

SG to help preclude the potential for

excessive leakage during all plant conditions.

The T/S limits imposed on primary-tosecondary leakage at operating conditions are

a maximum of 0.4 gpm (600 gpd) for all SGs

with a maximum of 150 gpd allowed for any

one SG.

The RG 1.121 criteria for establishing

operational leakage rate limits that require
40024 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

plant shutdown are based upon leak-beforebreak (LBB) considerations to detect a free

span crack before potential tube rupture

during faulted plant conditions. The 150 gpd

limit should provide for leakage detection

and plant shutdown in the event of the

occurrence of an unexpected single crack

resulting in leakage that is associated with

the longest permissible crack length.

Regulatory Guide 1.121 acceptance criteria

for establishing operating leakage limits are

based on LBB considerations such that plant

shutdown is initiated if the leakage

associated with the longest permissible crack

is exceeded. The longest permissible crack is

the length that provides a factor of safety of

1.43 against bursting at faulted conditions

maximum pressure differential. A voltage

amplitude of 8.8 volts for typical ODSCC

corresponds to meeting this tube burst

requirement at a lower 95% prediction limit

on the burst correlation coupled with 95/95

lower tolerance limit material properties.

Alternate crack morphologies can correspond

to 8.8 volts so that a unique crack length is

not defined by the burst pressure versus

voltage correlation. Consequently, typical

burst pressure versus through-wall crack

length correlations were used to define the

‘‘longest permissible crack’’ for evaluating

operating leakage limits. Consistent with the

cycle 13, 14 and 15 license amendment

requests for plugging criteria, and Section 5

of Enclosure 1 of the GL, operational leakage

limits will remain at 150 gpd per SG. Axial

cracks leaking at this level are expected to

provide LBB protection at both the SLB

pressure differential of 2560 psi and, while

not part of any established LBB methodology,

LBB protection will also be provided at a

value of 1.43 times the SLB pressure

differential. Thus, the 150 gpd limit provides

for plant shutdown prior to reaching critical

crack lengths for SLB conditions.

Additionally, this LBB evaluation assumes

that the entire crevice area is uncovered

during blowdown. Partial uncovery will

provide benefit to the burst capacity of the

intersection.

3) The proposed license amendment does

not involve a significant reduction in margin

of safety.

The use of the voltage-based bobbin probe

interim TSP elevation plugging criteria at

Cook Nuclear Plant Unit 1 is demonstrated to

maintain SG tube integrity commensurate

with the criteria of RG 1.121. Regulatory

Guide 1.121 describes a method acceptable to

the NRC staff for meeting GDC [General

Design Criteria] 14, 15, 31, and 32 by

reducing the probability or the consequences

of SG tube rupture. This is accomplished by

determining the limiting conditions of

degradation of SG tubing, as established by

in-service inspection, for which tubes with

unacceptable cracking should be removed

from service. Upon implementation of the

criteria, even under the worst case

conditions, the occurrence of ODSCC at the

TSP elevations is not expected to lead to a

SG tube rupture event during normal or

faulted plant conditions. It will be confirmed

by analysis and calculation that EOC

distribution of crack indications at the TSP

elevations will result in acceptable primaryto-secondary leakage during all plant

conditions and that radiological

consequences are not adversely impacted.

In addressing the combined effects of a

LOCA [loss-of-coolant accident] and SSE

[safe-shutdown earthquake] on the SG

component (as required by GDC 2), it has

been determined that tube collapse may

occur in the SGs at some plants. The

postulated tube collapse results from a

deformation of TSPs as a result of lateral

loads at the wedge supports at the periphery

of the plate. The lateral loads result from the

combined effects of the LOCA rarefaction

wave and SSE loadings. The resulting

pressure differential on the deformed tubes

may then cause some of the tubes to collapse.

There are two issues associated with a

postulated SG tube collapse. First, the

collapse of SG tubing reduces the RCS

[reactor coolant system] flow area through

the tubes. The reduction in flow area

increases the resistance to flow of steam from

the core during a LOCA which, in turn, may

potentially increase peak clad temperature.

Second, there is a potential that partial

through-wall cracks in tubes could progress

to through-wall cracks during tube

deformation or collapse.

Consequently, since the LBB methodology

is applicable to the Cook Nuclear Plant Unit

1 reactor coolant loop piping, the probability

of breaks in the primary loop piping is

sufficiently low that they need not be

considered in the structural design of the

plant. The limiting LOCA event becomes

either the accumulator line break or the

pressurizer surge line break. Loss of coolant

accident loads for the primary pipe breaks

were used to bound the Cook Nuclear Plant

Unit 1 smaller breaks. The results of the

analysis using the larger break inputs show

that the LOCA loads were found to be of

insufficient magnitude to result in SG tube

collapse or significant deformation.

Addressing RG 1.83 [In-Service Inspection

of PWR Steam Generator Tubes]

considerations, implementation of the bobbin

coil probe, voltage-based interim tube

plugging criteria of 2 volts is supplemented

by enhanced eddy current inspection

guidelines to provide consistency in voltage

normalization, a 100% eddy current

inspection sample size at the TSP elevation

per T/S, and MRPC [motorized RPC]

inspection requirements for the larger

indications left in-service to characterize the

principal degradation as ODSCC.

As noted previously, implementation of

the TSP elevation plugging criteria will

decrease the number of tubes which must be

repaired. The installation of SG tube plugs

reduces the RCS flow margin. Thus,

implementation of the plugging criteria will

maintain the margin of flow that would

otherwise be reduced in the event of

increased tube plugging.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a significant reduction in margin

with respect to plant safety as defined in the

FSAR or any Bases of the plant T/Ss.

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: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Mark Reinhart,

Acting

Northern States Power Company,

Docket No. 50-282, Prairie Island

Nuclear Generating Plant, Unit No. 1,

Goodhue County, Minnesota

Date of amendment request: July 15,

1996

Description of amendment request:

The proposed amendment would allow

the use of the moveable incore detector

system for measurement of the core

peaking factors with less than 75% and

greater than or equal to 50% of the

detector thimbles available. The

amendment request is a one-time only

change for Prairie Island, Unit 1,

Operating Cycle 18. It is being

submitted to allow for continued

operation if the number of detector

thimbles drops below 75%.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes do not involve an

increase in the probability of an accident

previously evaluated. The moveable incore

detector system is used only to provide

confirmatory information on the neutron flux

distribution and is not required for the daily

safe operation of the core. The system is not

a process variable that is an initial condition

in the accident analyses. The only accident

that the moveable incore detector system

could be involved in is the breaching of the

detector thimbles which would be enveloped

by the small break loss of coolant accident

(LOCA) analysis. As the proposed changes do

not involve any changes to the system’s

equipment and no equipment is operated in

a new or more harmful manner, there is no

increase in the probability of such an

accident.

The proposed amendments would not

involve an increase in the consequences of an

accident previously evaluated. The moveable

incore detector system provides a monitoring

function that is not used for accident

mitigation (the system is not used in the

primary success path for mitigation of a

design basis accident). The ability of the

reactor protection system or engineered
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40025

safety features system instrumentation to

mitigate the consequences of an accident will

not be impaired by the proposed changes.

The small break LOCA analysis (and thus its

consequences) continues to bound potential

breaching of the system’s detector thimbles.

With greater than or equal to 50% and less

than 75% of the detector thimbles available,

core peaking factor measurement

uncertainties will be increased, which could

impact the core peaking factors and as a

result could affect the consequences of

certain accidents. However, any changes in

the core peaking factors resulting from

increased measurement uncertainties will be

compensated for by conservative

measurement uncertainty adjustments in the

Technical Specifications to ensure that

pertinent core design parameters are

maintained. Sufficient additional penalty is

added to the power distribution

measurements such that this change will not

impact the consequences of any accident

previously evaluated.

Therefore, based on the conclusions of the

above analysis, the proposed changes will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed amendments would not

create the possibility of a new or different

kind of accident previously evaluated as they

only affect the minimum complement of

equipment necessary for operability of the

moveable incore detector system. There is no

change in plant configuration, equipment or

equipment design. No equipment is operated

in a new manner. Thus the changes will not

create any new or different accident causal

mechanisms. The accident analysis in the

Updated Safety Analysis Report remains

bounding.

Therefore, based on the conclusions of the

above analysis, the proposed changes will

not create the possibility of a new or different

kind of accident.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

The proposed changes will not involve a

significant reduction in a margin of safety.

The reduction in the minimum complement

of equipment necessary for the operability of

the moveable incore detector system could

only impact the monitoring/calibration

functions of the system. Reduction of the

number of available moveable incore detector

thimbles to the 50% level does not

significantly degrade the ability of the system

to measure core power distributions. With

greater than or equal to 50% and less than

75% of the detector thimbles available, core

peaking factor measurement uncertainties

will be increased, but will be compensated

for by conservative measurement uncertainty

adjustments in the Technical Specifications

to ensure that pertinent core design

parameters are maintained. Sufficient

additional penalty is added to the power

distribution measurements such that this

change does not impact the safety margins

which currently exist. Also, the reduction of

available detector thimbles has negligible

impact on the quadrant power tilt and core

average axial power shape measurements.

Sufficient detector thimbles will be available

to ensure that no quadrant will be

unmonitored.

Based on these factors, the proposed

changes in this license amendment will not

result in a significant reduction in the plant’s

margin of safety, as the core will continue to

be adequately monitored.

Based on the evaluation above, and

pursuant to 10 CFR 50, Section 50.91,

Northern States Power Company has

determined that operation of the Prairie

Island Nuclear Generating Plant in

accordance with the proposed license

amendment request does not involve any

significant hazards considerations as defined

by NRC regulations in 10 CFR 50, Section

50.92.

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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: Mark Reinhart,

Acting

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 31,

1996

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to add

a Limiting Condition for Operation

(LCO) for trisodium phosphate (TSP)

and increase the minimum required

amount of TSP contained in the

containment sump mesh baskets.

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.

Trisodium Phosphate Dodecahydrate (TSP)

is stored in the containment sump to raise

the pH of the sump and spray water

following a loss of coolant accident (LOCA).

As the pH of the water increases, more

radioactive iodine is kept in solution and the

possibility of airborne radioactivity leakage is

decreased. An additional advantage of a

higher pH is the beneficial reduction in

chloride stress corrosion cracking (SCC) of

austenitic stainless steel components in the

containment following a LOCA.

This chemical is an accident mitigator, not

an accident initiator in that it is not used

until after an accident (i.e., a LOCA) has

occurred. At the time it begins to go into

solution, the accident has occurred,

containment spray has been activated and

water is collecting in the containment sump.

Therefore, increasing the Technical

Specification (TS) minimum amount of TSP

verified to be in containment will not involve

a significant increase of the probability of an

accident previously evaluated.

The Updated Safety Analysis Report

(USAR), Section 14.15, ‘‘Loss of Coolant

Accident,’’ does not take credit for a postLOCA minimum containment sump pH

adjustment to 7.0 for the iodine removal and

retention calculation until ten hours after

initiation of the event. Increasing the amount

of TSP (based on recent re-analysis) in the

containment sump ensures that a pH greater

than or equal to 7.0 is achieved and therefore

does not increase the consequences of any

accident previously evaluated.

The proposed change to TS 2.3(4)

represents a new Limiting Condition for

Operation (LCO) which is added to establish

overall consistency with the CE STS

[Combustion Engineering Standard Technical

Specifications] for TSP requirements. The

proposed change establishes a minimum TSP

volume that must be maintained during

operating Modes 1 and 2 to ensure that a pH

greater than or equal to 7.0 is achieved

within four hours following a LOCA; as well

as, establishing times for accomplishing

corrective actions should the LCO not be met.

Therefore, this change does not significantly

increase the probability or consequences of

any accident previously evaluated.

The proposed change to TS 3.6(2)d(i)

revises the required surveillance inventory of

the TSP baskets consistent with the

aforementioned calculation to ensure that a

pH greater than or equal to 7.0 is achieved.

Therefore, this change does not increase the

consequences of any accident previously

evaluated.

The proposed change to TS 3.6(2)d(ii)

moves the surveillance test amounts of

chemical and water used from the

Specification to the Basis section. This

relocation will not alter the test method or

acceptance criteria.

In the Basis, the amount of TSP used in the

test is changed to reflect the ratio of TSP to

water that would be found in the

containment sump following a LOCA. The

specified concentration of boron in the test

reflects the highest concentration that could

be found in the containment sump following

a LOCA. The test temperature is changed to

115 - 125°F, which is well below the

temperature expected to be found in the

containment sump following a LOCA. The

decanting of the solution does not change the

intent of the test method since the dissolving

period will still be conducted without

agitation. Therefore, these changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.
40026 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

TSP is currently present in the

containment sump. The addition of TSP

ensures that a pH greater than or equal to 7.0

is achieved following a LOCA. The increase

in TSP inventory will be accomplished via a

modification to be installed during the 1996

Refueling Outage.

The proposed change to TS 2.3(4)

represents a new LCO which is added to

establish overall consistency with the CE STS

for TSP requirements. The proposed change

establishes a minimum TSP volume that

must be maintained during operating Modes

1 and 2 to ensure that a pH greater than or

equal to 7.0 is achieved following a LOCA,

as well as, establishing corrective action term

limits should the LCO not be met. This

proposed change does not create a possibility

of a new or different kind of accident from

any previously analyzed.

The proposed change to TS 3.6(2)d(ii)

moves the surveillance test amounts of

chemical and water used from the

Specification to the Basis section to be

consistent with the CE STS. This relocation

will not alter the test method or acceptance

criteria. In the Basis section, the amount of

TSP used in the test is changed to reflect the

ratio of TSP to water that would be found in

the containment following a LOCA. The

specified concentration of boron in the test

reflects the highest concentration that could

be found in the containment sump following

a LOCA. The test temperature is changed to

a range of 115 - 125°F which is well below

the temperature expected to be found in the

containment sump following a LOCA. The

decanting of the solution does not change the

intent of the test method since the dissolving

period will still be conducted without

agitation. Therefore, these changes will 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.

TSP is stored in the containment lower

level to raise the pH of the containment sump

and recirculated spray water following a

LOCA. As the pH of the water increases,

more radioactive iodine is kept in solution

and the possibility of airborne radioactivity

leakage is decreased. Additionally, a higher

pH has the beneficial effect of reducing the

possibility of chloride stress corrosion

cracking of austenitic stainless steel

components in the containment.

The proposed change to TS 2.3(4)

represents addition of a new LCO for TSP

requirements during power operations and

hot standby consistent with CE STS. This

change does not involve a significant

reduction in a margin of safety.

TS 3.6(2)d(i) requires verification that a

minimum volume of TSP is contained in the

storage baskets in containment. This change

proposes to increase that volume consistent

with the latest ABB/CE calculation. The

increased volume will ensure that the

containment sump, when filled with water

from the Reactor Coolant System, Safety

Injection Refueling Water Tank, Safety

Injection Tanks and Boric Acid Storage

Tanks, will have a pH greater than or equal

to 7.0 within four hours following a LOCA.

Therefore, this change does not involve a

reduction in a margin of safety.

The proposed change to TS 3.6(2)d(ii)

would move the surveillance test amounts of

chemical and water used from the

Specification to the Basis section. This

relocation is consistent with the CE STS and

will not alter the test method or acceptance

criteria. In the Basis, the amount of TSP used

in the test is changed to reflect the ratio of

TSP to water that would be found in the

containment following a LOCA. The

specified concentration of boron in the test

reflects the highest post-LOCA concentration

that could be found in the containment. The

test temperature is changed to a range of 115

- 125°F which is well below the temperature

expected to be found in the containment

sump following a LOCA. The decanting of

the solution does not change the intent of the

test method since the dissolving period will

still be conducted without agitation.

Therefore, these changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: July 15,

1996

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

allow the use of either zircaloy or

ZIRLO cladding and add a reference to

Westinghouse Topical Report, WCAP12610, June 1990.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed revision to TS 4.3.2 is based

on improved STS 4.2 of NUREG-1432. ZIRLO

is similar in chemical composition, physical

and mechanical properties to Zircaloy-4, but

features improved corrosion performance and

dimensional stability. These characteristics

ensure that fuel rod cladding integrity and

fuel assembly structural integrity are

maintained. Fuel assemblies manufactured

with ZIRLO clad fuel rods meet the same

design bases requirements as fuel assemblies

manufactured with Zircaloy-4 cladding and

the regulatory requirements of 10 CFR 50.46

are applicable to either material.

No concerns have been identified

pertaining to reactor operation with a core

comprised of fuel assemblies manufactured

with Zircaloy-4 clad rods and fuel assemblies

manufactured with ZIRLO clad rods. ZIRLO

clad fuel rods do not require a change to the

FCS [Fort Calhoun Station] reload design and

safety analysis limits. Radiological

consequences of previously evaluated

accidents are not increased because the safety

analysis dose predictions are not sensitive to

the type of cladding material used. The

proposed limited substitution of zirconium

alloy or stainless steel filler rods in

accordance with NRC-approved fuel rod

configurations will allow leaking fuel rods

(or potential leakers) to be removed.

Therefore, the radiological consequences of

accidents previously evaluated in the FCS

Updated Safety Analysis Report (USAR) are

not increased by this change.

The revisions to TS 4.3.2 listed above will

not result in a change to any of the process

variables that might initiate an accident or

affect the radiological release for an accident.

The operating limits will not be changed and

the analysis methods to demonstrate

operation within the limits will remain in

accordance with NRC-approved

methodology. There are no physical changes

to the plant associated with the change to TS

4.3.2 other than the changes to the fuel

assemblies. Therefore, this revision does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated because the safety

analysis to be performed for each cycle will

continue to demonstrate compliance with all

fuel safety design bases.

The proposed revision of TS 4.3.2 is

supported by Westinghouse Topical Report,

WCAP-12610, ‘‘VANTAGE + Fuel Assembly

Report,’’ dated June 1990 (Westinghouse

Proprietary). This topical report describes the

fuel rod design bases, criteria and models,

which are affected by the use of ZIRLO

cladding. Consequently, WCAP-12610 is

proposed for addition to the list of analytical

methods located in TS 5.9.5b that are used

to determine the core operating limits.

Based on the above discussion, these

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Fuel assemblies manufactured with ZIRLO

clad fuel rods must meet original design

criteria and thus they will not be an initiator

for any new or different kind of accident. All

design and performance criteria will

continue to be met by fuel assemblies

manufactured with ZIRLO clad fuel rods and
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40027

no new single failure mechanisms have been

found.

The use of fuel assemblies manufactured

with ZIRLO cladding does not involve any

alterations to plant equipment or procedures

that would introduce any new or unique

operational modes or accident precursors.

The substitution of zirconium alloy, stainless

steel filler rods, or lead test assemblies for

fuel rods will be limited to NRC-approved

fuel rod configurations. Therefore, the

possibility of a new or different kind of

accident from any accident previously

evaluated is not created by this change.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The use of fuel assemblies manufactured

with ZIRLO clad rods does not change the

proposed FCS reload design and safety

analysis limits. The normal operating

conditions allowed for in the Technical

Specifications will be taken into

consideration for the use of these fuel

assemblies. For each cycle reload core, the

fuel assemblies will be evaluated using NRCapproved reload design methods to include

consideration of the core physics analysis

peaking factors and core average linear heat

rate effects.

NRC-approved methods will also be used

to analyze each configuration of zirconium

alloy or stainless steel filler rods in fuel

assemblies to demonstrate continued safe

operation within the limits that assure

acceptable plant response to accidents and

transients. Therefore, this change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

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

3502

NRC Project Director: William H.

Bateman

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 amendment request: June 21,

1996

Description of amendment request:

The proposed amendment would

change the frequency of instrument

channel calibrations in Table 4.1-1,

‘‘Minimum Frequencies for Checks,

Calibrations and Test of Instrument

Channels’’ to accommodate operation

with a 24-month 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. Does operation with the proposed

license amendment involve a significant

increase in the probability or consequences

of any accident previously evaluated?

Response:

The proposed changes do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated. The proposed changes are being

made to extend the calibration frequency to

24-months for the:

Pressurizer Pressure; Accumulator Level

and Pressure; andVolume Control Tank

Level.

These changes are being made, using the

guidance of Generic Letter 91-04, to

accommodate a 24-month operating cycle.

The proposed changes in the calibration

frequencies do not involve any plant

hardware changes (other than alarm

adjustments) or the way the systems

function. The results of the instrumentation

drift analysis, loop accuracy/set point

calculations and the evaluation of channel

uncertainties indicate the calibrations can be

safely extended to accommodate the 24-

month operating cycle.

The four pressurizer pressure channels are

used for high and low pressure protection

(i.e., reactor trip and safety injection) and for

overpower-overtemperature protection. Three

of the pressure channels are also used for

pressure control and compensation signals

for rod control. Pressurizer pressure

indication is also provided in the control

room for use during normal operation and

while using the EOPs (emergency operating

procedure). The loop accuracy/setpoint

calculations confirm that sufficient margin

exists between the pressurizer high and low

pressure reactor trip, low pressurizer

pressure SI [safety injection], and

overtemperature delta-temperature analytical

limits and the existing field trip settings

based on an extended calibration interval. A

small increase in pressurizer pressure normal

indication uncertainty due to increased

sensor drift is within the readability of the

indicator and has been incorporated into the

pressurizer pressure initial conditions used

in the evaluation of channel uncertainties

(Reference 15) [see application dated June 21,

1996]. The post-accident indication

uncertainties remain bounded by the existing

uncertainties used in the EOPs. Assurance

that the RPS [reactor protection system] and

ESF [engineered safety feature]

instrumentation and protection logic relays

will function as required is also provided by

on-line surveillance (channel checks

performed each shift and quarterly channel

functional tests) that are designed to detect

potential instrument failures and verify

operability of pressurizer pressure channels.

Water level and pressure in each

accumulator is monitored by two redundant

channels designed to provide indication in

the control room. High and low level alarm

functions alert the operator to initiate

operations to maintain the accumulator water

volume or pressure within the Technical

Specifications limits. The level and pressure

instrumentation do not provide an active

protective or control function and are not

required to mitigate an accident condition.

The level (or volume) and pressure limits are

important since they are initial conditions

assumed in the safety analysis. The loop

accuracy/setpoint calculations for

accumulator level and pressure were updated

to include conservative values for 30-month

calibration uncertainties using Westinghouse

sensor drift values and extrapolated vendor

specified uncertainties for rack and

indicating components consistent with

industry methods. The increased indicator

uncertainty has been evaluated for both input

parameters (accumulator level and pressure)

assumed for the LOCA [loss-of-coolant

accident] and Containment Integrity events

(Reference 15) and a non significant increase

in both the peak clad temperature and

containment pressure was identified.

The volume control tank (VCT) level

instrumentation is not required to mitigate

the consequences of an accident. The

instrumentation provides control room

indication and initiates automatic actions of

the chemical and volume control system

(e.g., diverts letdown to the holdup tanks on

high level, initiates makeup on low level,

changes the charging pump suction on low

low level). The loop accuracy/setpoint

calculation for VCT level, updated based on

the increased drift and uncertainty,

determined that the existing setpoints remain

valid to ensure the VCT instrumentation can

perform the required design function.

2. Does operation with the proposed

license amendment create the possibility of

a new or different kind of accident from any

previously evaluated?

Response:

The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated. The

proposed changes extend the calibration

frequency to 24 months for the Pressurizer

Pressure, Accumulator Pressure and Level,

and Volume Control Tank Level

instrumentation to accommodate a 24-month

operating cycle. The proposed changes in

calibration frequencies do not involve any

plant hardware changes, nor do they change

the way that the systems function.

The extension of the calibration and

surveillance test intervals were evaluated and

the results, documented in Reference 15,

indicate that the calibrations can be safely

extended to accommodate the 24-month

operating cycle.

3. Does operation with the proposed

license amendment involve a significant

reduction in a margin of safety?

Response:

The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes extend the calibration

frequency to 24 months for the Pressurizer

Pressure, Accumulator Pressure and Level,

and Volume Control Tank Level

instrumentation to accommodate a 24-month

operating cycle.

The proposed changes result in an

increased instrument channel uncertainty for

the pressurizer pressure. An evaluation

(Reference 15) has determined that: all
40028 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

current cycle 9 safety analysis limits based

on pressurizer pressure uncertainties remain

bounding for extended surveillance intervals

(high and low pressure trips); the safety

analysis limits for K1 (a constant used in the

overtemperature [DELTA] T trip setpoint)

remain applicable; and, Engineered Safety

Feature Actuation System trip settings based

on pressurizer pressure uncertainty remain

bounding (low pressure safety injection).

The proposed changes result in an

increased instrument channel uncertainty for

the accumulator level and pressure. An

evaluation (Reference 15) has determined

that increasing the uncertainty results in nonsignificant (defined by 10 CFR 50.46(a)(3)(i)

as less than 50°F) increases in the total peak

clad temperature (less than 35°F) for the large

break and small break LOCA but the values

remain well within regulatory acceptance

criteria. The evaluation also determined that

the peak calculated pressure in containment

following a LOCA would increase due to the

lower bound on pressure and the higher

bound on volume in the accumulators. An

assessment of the approximate effect on the

peak containment pressure determined that

the Technical Specification integrated leak

rate testing value of 42.42 psig (the licensing

basis peak pressure) remains bounding.

The proposed changes result in an

increased instrument channel uncertainty for

the VCT level but there are no changes to any

margins of safety because this

instrumentation supports a control function.

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: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

Southern California Edison Company,

et al., Docket No. 50-206, San Onofre

Nuclear Generating Station, Unit No. 1,

San Diego County, California

Date of amendment request:

December 22, 1995

Description of amendment request:

The proposed change would revise the

San Onofre Unit 1 License Condition to

delete a reference to License Condition

2.C(4) from License Condition 2.D. This

change is being requested to eliminate a

reporting requirement for violations of

the physical protection plans that is

redundant to reporting requirements in

10 CFR 73.71 and 10 CFR 73 Appendix

G.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility according

to this proposed change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. The proposed change is considered an

administrative change. It has no impact on

the probability or consequences of any of the

accidents previously evaluated. This change

revises License Condition 2.D to remove the

burden of duplicate reporting requirements.

This change does not affect the physical

protection program as previously approved

by the Nuclear Regulatory Commission

(NRC).

A reporting requirement in License

Condition 2.D is being revised to remove the

reference to License Condition 2.C(4) for the

physical protection program. The reporting

requirements for the physical protection

program are located in the regulations, 10

CFR 73.71 and 10 CFR 73 Appendix G.

Therefore, the probability and

consequences of an accidently previously

evaluated are not affected by these proposed

changes.

2. Will operation of the facility according

to this proposed change create the possibility

of a new or different kind of accident from

any accident previously evaluated.

No. This proposed change is considered an

administrative change. It has no impact on

equipment, systems, or structures such that

a new or different kind of accident is created.

This change revises License Condition 2.D to

remove duplicate and unnecessary reporting

requirements for the physical protection

program. There is no change associated with

the implementation and maintenance of the

physical protection program as previously

approved by the NRC.

Therefore, the possibility of a new or

different kind of accident from an accident

previously evaluated is not created.

3. Will operation of the facility according

to this proposed change involve a significant

reduction in a margin of safety?

No. This proposed change is considered an

administrative change only. It has no impact

on the margin of safety associated with the

physical protection program. This change

revises License Condition 2.D to remove

duplicative and unnecessary reporting

requirements for the physical protection

program. The maintenance and

implementation of the physical protection

program is not affected by this change.

Therefore, there will not be a significant

reduction in a margin of safety.

The NRC staff has reviewed the

analysis of the licensee 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: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713

Attorney for licensee: James A.

Beoletto, Esquire, Southern California

Edison Company, P.O. Box 800,

Rosemead, California 91770

NRC Project Director: Seymour H.

Weiss

Southern California Edison Company,

et al., Docket No. 50-206, San

OnofreNuclear Generating Station, Unit

No. 1, San Diego County, California

Date of amendment request: March

13, 1996

Description of amendment request:

The proposed change would revise San

Onofre Unit 1 License Condition 2.D in

the Operating (Possession Only) License

to remove a reporting requirement that

is redundant to reporting requirements

in 10 CFR 50.72 and 50.73.

Additionally, the proposed change

would make administrative and

editorial changes in the Permanently

Defueled Technical Specifications,

which constitute Appendix A of the

Operating (Possession Only) License.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility according

to this proposed change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. San Onofre Nuclear Generating

Station, Unit 1 (SONGS 1) has been

permanently shut down with its reactor

defueled and spent fuel from the reactor

stored in the spent fuel pool. The proposed

change will not modify any of the existing

plant configurations, controls, procedures, or

Permanently Defueled Technical

Specifications (PDTS) requirements

necessary to assure the integrity and safe

operation of the spent fuel pool.

The requested change to License Condition

2.D will result in not requiring violations of

the PDTS to be reported based on License

Condition 2.D. The basis for this change is

that all types of reportable events applicable

to a defueled plant are covered by 10 CFR

50.72 and 50.73, which SONGS 1 is required

to implement. Any other reporting

requirements imposed through a license

condition are redundant to reporting

requirements contained in 10 CFR 50.72 and

50.73. Therefore, this change is

administrative.

The requested changes to the PDTS are also

administrative in nature. They consist of

changes to reflect the current nuclear

organization and responsibilities, modify

administrative requirements relating to the

Onsite Review Committee, modify a

requirement relating to Final Safety Analysis

Report documentation using NRC guidance,

and make editorial corrections and

improvements in the text. Since these

changes are administrative, they have no

effect on the accidents previously evaluated.
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40029

Therefore, operation of the facility in

accordance with this proposed change will

not involve a significant increase in the

probability or consequences of an accidently

previously evaluated.

2. Will operation of the facility according

to this proposed change create the possibility

of a new or different kind of accident from

any accident previously evaluated.

No. The proposed changes do not alter the

design, configuration, or method of operation

of the plant. The changes to License

Condition 2.D and the PDTS are

administrative or editorial.

Therefore, operation of the facility in

accordance with this proposed change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will operation of the facility according

to this proposed change involve a significant

reduction in a margin of safety?

No. The proposed changes do not alter the

design, configuration, or method of operation

of the plant. Since the proposed changes are

administrative or editorial, the existing plant

safety margins are not reduced.

Therefore, operation of the facility in

accordance with this proposed change will

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

analysis of the licensee 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: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713

Attorney for licensee: James A.

Beoletto, Esquire, Southern California

Edison Company, P.O. Box 800,

Rosemead, California 91770

NRC Project Director: Seymour H.

Weiss

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of amendment requests: May 29,

1996

Description of amendment requests:

The licensee proposes to revise

improved Technical Specification (TS)

3.5.1, ‘‘Safety Injection Tanks (SITs),’’ to

increase the minimum boron

concentration in the safety injection

tanks from 1850 parts per million (ppm)

to 2200 ppm. This TS change is being

requested to support the planned

increase in the operating cycle length.

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.

Southern California Edison (Edison) is

increasing the minimum boron concentration

to maintain the ability of the Safety Injection

Tanks (SITs) to perform their intended safety

function consistent with the increase in fuel

enrichment up to 4.8 weight percent (w/o)

Uranium-235 and changing the burnable

poison from B4C to Erbia (Erbium-Oxide

Er2O3 and fuel mixture) to increase the length

of the operating cycle. Increasing the

minimum boron concentration in the SITs

will maintain the ability of the Emergency

Core Cooling System (ECCS) to control core

reactivity during and following an accident.

No change is being made to the design of

the safety injection system. Consequently,

there will be no impact on the probability of

initiating an accident which has been

previously evaluated.

Increasing the boron concentration in the

SITs will ensure the ability of this system to

mitigate the accidents for which it is

required. No other accident conditions,

design conditions, Technical Specifications,

or Technical Specification Bases are affected

by this proposed change in boron

concentration.

Therefore, the operation of the facility in

accordance with this proposed change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There is no change in plant design or

operational methodology imposed by the

increase in SIT boron concentration. This

increase in boron concentration is required

because Edison is increasing the fuel

enrichment up to 4.8 w/o Uranium-235 and

changing the burnable poison from B4C to

Erbia to achieve a longer cycle length.

Therefore, additional negative reactivity is

required at the beginning of the fuel cycle for

these alternate coolant sources.

Edison believes this change in the SIT

minimum boron concentration limit is, in

essence, an administrative change. The SITs

are filled from the refueling water storage

tank (RWST), which has a technical

specification minimum boron concentration

requirement of 2350 ppm. Edison maintains

the RWST boron concentration higher than

the minimum limit. As a result, for the past

several years the SIT boron concentration has

been approximately 2500 ppm, even though

the technical specification lower limit is

1850 ppm. The maximum boron

concentration limit is not being changed.

Increasing the SIT minimum boron

concentration limit of the technical

specification narrows the existing operating

band, and maintaining the boron

concentration between 2200 ppm and 2800

ppm will keep the boron concentration

between the current band of 1850 ppm to

2800 ppm. Therefore, changing the SIT

minimum boron concentration from 1850

ppm to 2200 ppm does not involve a physical

change to the plant.

Therefore, the operation of the facility in

accordance with this proposed change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

With the increase in fuel enrichment up to

4.8 w/o Uranium-235 and changing the

burnable poison from B4C to Erbia to increase

the length of the operating cycle, increasing

the minimum boron concentration in the

SITs is required to maintain the current

margins of safety.

The calculations were performed to ensure

the core remains subcritical (i.e.,

conservatively 1% shutdown) with the

proposed boron concentration. In addition to

the conservative assumptions used in the

calculation, 50 ppm was added to the results.

Therefore, the operation of the facility in

accordance with this proposed change does

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770

NRC Project Director: William H.

Bateman

Southern Nuclear Operating Company,

Inc., Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request: June 12,

1996

Description of amendments request:

The proposed amendments would

revise the reactor core safety limits,

Overtemperature delta T (OTDT) and

Overpressure delta T (OPDT) reactor

trip setpoints and allowable values, and

the power distribution limits associated

with implementation of Relaxed Axial

Offset Control (RAOC) and FQ

surveillance. The proposed amendments

also include changes to the Bases

associated with these specifications and

surveillances.

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:
40030 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

1. The proposed safety limits, reactor trip

setpoints, HNF [high neutron flux] setpoints

for MSSVs [main steamline safety valves] out

of service, F[delta]H for LOPAR [low

parasitic], and RAOC strategy changes do not

increase the probability or consequences of

an accident previously evaluated in the

FSAR [Final Safety Analysis Report]. The

core safety limits and trip setpoints were

determined using the NRC reviewed and

approved DNB [departure from nucleate

boiling] methodologies, namely RTDP, and

approved DNB correlations. No new

performance requirements are being imposed

on any system or component in order to

support the revised core limits. Overall plant

integrity is not reduced. The DNB sensitive

transients that are protected by [OPDT] and

[OTDT] were reanalyzed or evaluated. The

DNB design criterion continues to be met.

None of these changes directly initiate an

accident; therefore, the probability of an

accident has not increased. No new

performance requirements are imposed on

any safety-related equipment. The acceptance

criteria for the reanalyses continue to be met;

therefore, the consequences of accidents

previously evaluated in the FSAR are not

significantly changed. All dose consequences

have been evaluated for these changes and all

acceptance limits continue to be met. All

safety analyses that use the revised [OTDT]

and [OPDT] setpoints continue to meet all

acceptance criteria. [Loss-of-coolant accident]

LOCA analyses are not affected by any of

these proposed changes.

2. The proposed Technical Specifications

changes do not create the possibility of a new

or different kind of accident than any

accident already evaluated in the FSAR. No

new accident scenarios, failure mechanisms

or limiting single failures are introduced as

a result of the proposed changes. The

proposed Technical Specifications changes

have no adverse effects on any safety-related

system and do not challenge the performance

or integrity of any safety-related system. The

DNB design criterion continues to be met.

The use of the revised core limits, reactor trip

setpoints and RAOC have been shown to

allow FNP [Farley Nuclear Plant] to operate

in a safe configuration. Therefore, the

possibility of a new or different kind of

accident is not created.

3. The proposed Technical Specifications

changes do not involve a significant

reduction in a margin of safety. All accident

analysis acceptance criteria continue to be

met. The DNB design criterion remains

unchanged. The DNBR [departure from

nucleate boiling ratio] design limit values

have not changed. Therefore, the DNB design

limit values associated with the DNB

methodology and correlations, upon which

the Technical Specifications changes are

based, do not result in a significant reduction

in the margin of safety because the DNB

design criterion continues to be met. The

proposed revisions to the Technical

Specifications result in an operating

configuration consistent with the analytic

assumptions (including LOCA analyses) used

to form the bases of 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: 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

Southern Nuclear Operating Company,

Inc., Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request: June 20,

1996

Description of amendments request:

The proposed amendments would

revise the Technical Specifications (TS)

to incorporate the requirements of 10

CFR Part 50, Appendix J, Option B. The

Administrative Controls portion would

be revised to establish and reference a

‘‘Containment Leakage Rate Testing

Program’’ in accordance with the NRC’s

Regulatory Guide 1.163 dated

September 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability of

consequences of an accident previously

evaluated. The proposed changes provide a

mechanism within the TS for implementing

a performance-based leakage rate test

program which was promulgated by the

revision to 10 CFR [Part] 50 to incorporate

Option B to Appendix J. The proposed

changes do not involve any physical or

operational changes to structures, systems or

components. The proposed TS Limiting

Conditions for Operation (LCO) are

consistent with 10 CFR [Part] 50, Appendix

J requirements and are equivalent to the

current LCO requirements. The current safety

analyses and safety design basis for the

accident mitigation functions of the

containment, the airlocks, and the

containment isolation valves are maintained.

Since the allowable containment leakage is

still maintained within the analyzed limit

assumed in the accident analyses, there is no

adverse effect on either onsite or offsite dose

consequences. Furthermore, containment

leakage is not an accident initiator.

Therefore, these changes will not increase the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

analyzed. The proposed changes do not

involve any physical or operational changes

to structures, systems or components. No

new failure mechanisms beyond those

already considered in the current plant safety

analyses are introduced. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

Extending Type A, B, and C test intervals

from those currently provided in the TS to

those provided for in 10 CFR [Part] 50

Appendix J, Option B slightly increases risk

due to an increased likelihood of

containment leakage corresponding to the

increased testing intervals. However, this is

somewhat compensated by the corresponding

risk reduction benefits received from the

reduction in component cycling, stress, and

wear associated with the increased intervals.

When considering the total integrated risk,

which includes all analyzed accident

sequences, the additional risk associated

with increasing test intervals is negligible.

The NRC letter to NEI [Nuclear Energy

Institute] dated November 2, 1995, recognizes

that changes similar to the proposed changes

at FNP [Farley Nuclear Plant] are required to

implement Option B of 10 CFR [Part] 50,

Appendix J. In NUREG-1493, ‘‘PerformanceBased Containment Leak-Test Program,’’

dated September 1995, which forms the basis

for the Appendix J revision, the NRC

concludes that adoption of performancebased test intervals for Appendix J testing

will not significantly reduce the margin of

safety. The containment leak rate data and

component performance history at FNP are

consistent with the conclusions reached in

NUREG-1493 and NEI 94-01. Thus, the

proposed license amendments do not involve

a significant reduction in a margin of safety

and will continue to support the regulatory

goal of ensuring an essentially leak-tight

containment boundary.

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
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40031

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: May 28,

1996

Description of amendment request:

The proposed amendment would

increase the test interval for Technical

Specification (TS) 3/4.3.1.1, Reactor

Protection System Instrumentation from

monthly on a staggered test basis to

semiannually on a staggered test basis

for the control rod drive trip breakers

and the reactor trip module logic.

Additionally, the proposed amendment

would increase the test interval from

monthly to semiannually for the output

logic of the anticipatory reactor trip

system (ARTS) instrumentation as

specified in TS 3/4.3.2.3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

(1)

Operation of the DBNPS in

accordance with the proposed license

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Increasing the surveillance interval

will not affect the probability or

consequences of an accident previously

evaluated since performance of the

surveillance test only ensures

operability of the particular trip

function at the time of the test. The

licensee evaluated the maintenance

history and surveillance test results of

the control rod drive trip breakers,

reactor trip module logic, and ARTS

output logic to show these components

have consistently met their design and

operational requirements over the past 8

years.

(2) Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes do not modify

or affect system design, function,

operation, or manner of testing.

(3) Involve a significant reduction in

a margin of safety.

The licensee has performed a

reliability evaluation that indicates

insignificant change in reactor trip

system unavailability and a reduction in

the potential for spurious trips resulting

from testing which support the

conclusion that a significant reduction

in a margin of safety will not occur.

Based on the NRC staff review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: June 28,

1996

Description of amendment request:

The proposed amendment would revise

the Technical Specifications for

shutdown margin to allow calculational

determination of the highest worth

control rod. Editorial changes are also

included.

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) During refueling, maintenance may be

performed on either the control rods or the

control rod drive mechanisms. Controls, such

as refueling interlocks, are provided to assure

inadvertent criticality does not occur during

this maintenance. There are no proposed

revisions to these controls except to lower

the threshold for applicability, which

constitutes a more restrictive change.

These controls also continue to assure that

the new, higher minimum shutdown margin

is maintained to ensure the reactor can be

returned to a subcritical condition should an

inadvertent criticality occur. The proposed

alternate calculational method for highest

worth control rod has additional

conservatism to account for any uncertainties

in the calculation and provides equivalent

margin. Therefore, this change will not

significantly increase the probability or

consequences of any previously analyzed

accident.

(2) The proposed change does not

necessitate a physical alteration of the plant

in that no new or different type of equipment

will be installed. The proposed change does

propose a higher minimum shutdown margin

and a lower threshold of applicability for

CRD [control rod drive] maintenance, both of

which are more restrictive. The proposed

change will provide effective methods to

preserve the safety functions associated with

the prevention or automatic mitigation of

design basis accidents. Thus, this change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) The proposed changes to the controls

provided to allow control rod withdrawal for

the purposes of maintenance are more

restrictive and thus preserve the safety

functions associated with the prevention or

automatic mitigation of design basis

accidents. The addition of a higher minimum

shutdown margin requirement and the

proposed calculational alternative for highest

worth rod, does not decrease any of the safety

controls or functions to prevent inadvertent

criticalities and provides equivalent or higher

margins. Therefore, this change will not

significantly reduce a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

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

Ropes and Gray, One International

Place, Boston, MA 02110-2624

NRC Project Director: Jocelyn A.

Mitchell, Acting Directorboro, VT 05301

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: July 3,

1996

Description of amendment request:

The proposed amendment would

modify Kewaunee Nuclear Power Plant

(KNPP) Technical Specification (TS)

Section 4.2.b, ‘‘Steam Generator Tubes,’’

to: revise the plugging criteria for tubes

in the tubesheet crevice region; add new

inspection criteria for tubes evaluated

using the new plugging criteria; add

definitions of terms used in the new

plugging criteria; and add reporting

requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

1. Operation of the KNPP in accordance

with the proposed license amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The revised plugging criteria ensure that

tubes in the tubesheet with indication(s) are

sufficiently inspected and evaluated and, if

necessary, rolled to meet the proposed
40032 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

acceptance criteria based on the new

definitions of acceptable distance between

the indication and the rolled area. With

sufficient distance between the indication(s)

and the hard rolled region of the tube in the

tubesheet, tube rupture probability and the

consequences of tube rupture are the same as

previously analyzed. Additionally, the

potential for leakage is within previously

analyzed limits.

2. The proposed license amendment

request does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

Implementation of the proposed tube

plugging criteria and proposed inspection

acceptance criteria based on the proposed

definitions does not introduce any significant

changes to the plant design basis. Use of

these criteria will not introduce a mechanism

that will result in an accident initiated

outside of the tubesheet crevice region. Any

hypothetical accident as a result of tube

indications in the tubesheet crevice region of

the tube will be bounded by the existing tube

rupture analysis. Therefore, application of

the revised acceptance criteria for

indication(s) within the tubesheet crevice

region will not create the possibility of a new

or different kind of accident.

3. The proposed license amendment does

not involve a significant reduction in the

margin of safety.

The use of the proposed inspection criteria

and tube plugging acceptance criteria will

maintain the integrity of the tube bundle

commensurate with the requirements of

Regulatory Guide 1.121 under normal and

postulated accident conditions. The safety

factors used in verification of the strength of

tube(s) evaluated under the new plugging

criteria are consistent with the safety factors

in the ASME Boiler and Pressure Vessel Code

used for steam generator design. The leak

testing acceptance criteria are based on the

primary-to-secondary leakage limits in the

TSs and the Updated Safety Analysis Report

accident analyses will be maintained.

Therefore, the proposed TS change will not

result in a significant reduction in the margin

of safety.

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

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed NoSignificant

Hazards Consideration

Determination,And Opportunity For A

Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Northeast Utilities Service Company,

Docket No. 50-336, Millstone Nuclear

Power Station, Unit No. 2, New London,

Connecticut

Date of amendment request: July 3,

1996Brief

Description of amendment request:

The proposed amendments would

provide a one-time change to Technical

Specification 3.9.1, ‘‘Refueling

Operations, Boron Concentration.’’ The

proposed change would remove the

requirement that the boron

concentration in all filled portions of

the Reactor Coolant System be

‘‘uniform.’’

Date of publication of individual

notice in Federal Register: July 11, 1996

(61 FR 36583)

Expiration date of individual notice:

August 12, 1996

Location Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut and the Wateford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendment:

June 3, 1996, as superseded by

application dated June 25, 1996

Brief description of amendment

request: The proposed amendment

would revise Technical Specifications

3.3.11, ‘‘Post Accident Monitoring

Instrumentation,’’ and 5.5.2.13, ‘‘Diesel

Fuel Oil Testing Program.’’ The

amendment would reinstate provisions

of the current San Onofre Nuclear

Generating Station, Unit Nos. 2 and 3

technical specifications that were

revised as part of Amendment Nos. 127

and 116. These amendments adopted

the recommendations of NUREG-1432,

‘‘Standard Technical Specifications

Combustion Engineering Plants.’’

Date of individual notice in Federal

Register: July 2, 1996 (61 FR 34452)

Expiration date of individual notice:

August 1, 1996

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

April 25, 1995

Brief description of amendment

request: The proposed amendment

would add a reactor water cleanup

system high blowdown containment

isolation trip function and associated

limiting condition for operation and

surveillance requirements.

Date of individual notice in Federal

Register: June 28, 1996 (61 FR 33777)

Expiration date of individual notice:

July 29, 1996

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

June 6, 1995, as supplemented by letter

dated April 22, 1996.

Brief description of amendment

request: The proposed amendment

would make administrative and

editorial changes to Section 6.0 of the

technical specifications for WNP-2.Date

of individual notice in Federal Register:

June 28, 1996 (61 FR 33779)

Expiration date of individual notice:

July 29, 1996

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

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
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40033

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.

Carolina Power & Light Company,

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

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

March 29, 1996.

Brief description of amendment: The

amendment revises the technical

specifications (TS) to add an allowance

to complete a TS-required surveillance

within 24 hours of discovery of a missed

surveillance in accordance with the

guidance of Generic Letter (GL) 87-09,

‘‘Sections 3.0 and 4.0 of the Standard

Technical Specifications (STS) on the

Applicability of Limiting Conditions for

Operation and Surveillance

Requirements.’’

Date of issuance: July 8, 1996

Effective date: July 8, 1996

Amendment No. 170

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25669)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 8, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

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:

November 2, 1994

Brief description of amendments: The

amendments delete the content of

Appendix B, ‘‘Environmental Protection

Plan (EPP) (Nonradiological),’’ and

modify License Condition 2.C.(2) to

delete that portion which refers to the

EPP.

Date of issuance: July 8, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 149 and 143

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

the Environmental Protection Plan and

License Conditions.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25702)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 8, 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:

April 18, 1996

Brief description of amendment: The

amendment deleted a restriction on the

24-hour emergency diesel generator

operation test in Surveillance

Requirement 3.8.1.14 of the Technical

Specifications for the Grand Gulf

Nuclear Station, Unit 1. The deletion

allows the test to also be conducted

during power operation (i.e., during

Modes 1 and 2), instead of the current

requirement to only conduct the test

when the plant is shut down.

Date of issuance: July 15, 1996

Effective date: July 15, 1996

Amendment No: 124

Facility Operating License No. NPF29: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20847) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 15, 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.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Entergy Mississippi, Inc., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of application for amendment:

May 6, 1996

Brief description of amendment: The

amendment reflects that the name of

Mississippi Power & Light Company

(MP&L) has been changed to Entergy

Mississippi, Inc. The amendment

revises Operating License No. NPF-29

and the Antitrust Conditions for the

Grand Gulf Nuclear Station, Unit 1

(GGNS) to (1) add the phrase ‘‘(now

renamed Entergy Mississippi, Inc.)’’, (2)

replace the name of Mississippi Power

& Light Company (MP&L) by the name

Entergy Mississippi, Inc., and (3)

replace a footnote by the statement:

‘‘Amendment 125 resulted in a name

change for Mississippi Power & Light

Company (MP&L) to Entergy

Mississippi, Inc.’’.

Date of issuance: July 16, 1996

Effective date: July 16, 1996

Amendment No: 125

Facility Operating License No. NPF29. Amendment revises the operating

license.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28613)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 16, 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.

Florida Power and Light Company, et

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

Unit No. 1, St. Lucie County, Florida

Date of application for amendments:

June 1, 1996

Brief description of amendments:

Revise Technical Specifications to

reflect reduced reactor coolant system
40034 Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices

flows resulting from increased

percentage of plugged steam generator

tubes.

Date of Issuance: July 9, 1996

Effective Date: July 9, 1996

Amendment Nos.: 145

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

the Technical Specifications.

Date of initial notice in Federal

Register: June 7, 1996 (61FR29140). The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 9, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

July 26, 1995

Brief description of amendments: The

amendments modify the Technical

Specifications to allow operation with

up to plus or minus 18 steps of rod

misalignment at or below 90 percent

power.

Date of issuance: July 12, 1996

Effective date: July 12, 1996

Amendment Nos. 186 and 180Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 13, 1995

(60FR47616) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 12, 1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

May 7, 1996 (TSCR 247)

Brief description of amendment: The

amendment adopts the provisions of the

Standard Technical Specifications,

NUREG-1433, Rev. 1 which clarify

surveillance requirement applicability

and allow a maximum period of 24

hours to complete a surveillance

requirement upon discovery that the

surveillance has been missed.

Date of Issuance: July 15, 1996

Effective date: July 15, 1996

Amendment No.: 185

Facility Operating License No. DPR16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28615).

The Commission’s related evaluation of

this amendment is contained in a Safety

Evaluation dated July 15, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

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:

December 21, 1995

Brief description of amendments: The

amendments modify the Peach Bottom

Atomic Power Station Units 2 and 3

Facility Operating Licenses to provide

for elimination of outdated or

superseded material regarding, among

other things, environmental monitoring

and modifications to the low pressure

coolant injection system, and for making

the FOLs for both units consistent.

Date of issuance: July 15, 1996

Effective date: Units 2 and 3, as of the

date of issuance, to be implemented

within 30 days.

Amendments Nos.: 215 and 220

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

revised the Facility Operating Licenses.

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10396)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 15, 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.

Public Service Electric & Gas Company,

Docket No. 50-311, Salem Nuclear

Generating Station, Unit No. 2, Salem

County, New Jersey

Date of application for amendment:

May 7, 1996, as supplemented June 14,

1996

Brief description of amendment: The

amendment made a one-time change to

Technical Specification 3/4.7.6,

‘‘Control Room Emergency Air

Conditioning System,’’ which permits

refueling of Unit 2 with the Control

Room Emergency Air Conditioning

System (CREACS) inoperable in Modes

5 and 6. This change will expire after

the completion of the Control Room and

CREACS upgrade, currently in progress,

and the restart and entry into Mode 4 of

Unit 2 from the current outage.

Date of issuance: July 10, 1996

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment No. 165

Facility Operating License No. DPR75: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25710)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 10, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

April 22, 1996, as supplemented June

12, 1996

Brief description of amendments: The

amendments change the Technical

Specifications to implement 10 CFR Part

50, Appendix J, Option B, for the Type

A test by referring to Regulatory Guide

1.163, ‘‘Performance Based Containment

Leakage-Test Program.’’

Date of issuance: July 11, 1996

Effective date: Both units, As of date

of issuance, to be implemented within

30 days.

Amendment Nos. 184 and 166

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20856) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 11, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

Southern Nuclear Operating Company,

Inc., Alabama Power Company, Docket

Nos. 50-348 and 50-364, Joseph M.

Farley Nuclear Plant, Units 1 and 2,

Houston County, Alabama

Date of application for amendments:

June 24, 1996

Brief description of amendments: The

amendments approve a unit cycle
Federal Register / Vol. 61, No. 148 / Wednesday, July 31, 1996 / Notices 40035

specific (Unit 1, Cycle 14 and Unit 2,

Cycle 11) Technical Specification

change to Note 4 of Table 4.3-1 that

permits continued operation of both

Farley units without performing the

required surveillance of the manual

safety injection input to the reactor trip

circuitry for the current operating cycle

until the next unit shutdown, following

which, this testing has to be performed

prior to entering Mode 2.

Date of issuance: July 19, 1996

Effective date: July 19, 1996

Amendment Nos.: 120 and 112

Facility Operating License Nos. NPF2 and NPF-8: The amendments revised

the Technical Specifications.Public

comments requested as to proposed no

significant hazards consideration: Yes.

(61 FR 34880 dated July 3, 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 August 2, 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 amendment.

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

July 19, 1996.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, P.O.

Box 1369, Dothan, Alabama

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

January 2, 1996, as supplemented by

letter dated April 12, 1996.

Brief description of amendment: The

amendment would revise TS 3.9.4 and

its associated Bases to allow the

containment personnel airlock doors to

be open during core alterations and

movement of irradiated fuel in

containment.

Date of issuance: July 15, 1996

Effective date: July 15, 1996, to be

implemented within 30 days of the date

of issuance.

Amendment No.: 114

Facility Operating License No. NPF30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5819). The April 12, 1996, supplemental

letter provided clarifying information

and did not change the original no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

July 15, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of application for amendment:

April 4, 1996

Brief description of amendment: The

amendment revises the Technical

Specifications regarding secondary

containment integrity including

addition of required actions in the event

secondary containment integrity is not

maintained when required. It also

requires surveillance of the secondary

containment isolation valves under the

licensee’s in-service testing program.

Date of issuance: July 10, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 147

Facility Operating License No. DPR28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20859) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 10, 1996No

significant hazards consideration

comments received: No

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Dated at Rockville, Maryland, this 24th day

of July 1996.

For the Nuclear Regulatory Commission

Steven A. Varga, Director,

Division of Reactor Projects - I/II Office of

Nuclear Reactor Regulation

[Doc. 96-19317 Filed 7-30-95; 8:45 am]

BILLING CODE 7590-01-F

SECURITIES AND EXCHANGE

COMMISSION

[Rel. No. IC–22097; File No. 812–9992]

Continental Assurance Company, et al.

July 25, 1996.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’ or ‘‘Commission’’).

ACTION: Notice of application for

exemptions under the Investment

Company Act of 1940 (‘‘1940 Act’’).

APPLICANTS: Continental Assurance

Company (‘‘CAC’’), Valley Forge Life

Insurance Company (‘‘VFL,’’ together

with CAC, the ‘‘Companies’’),

Continental Assurance Company

Variable Life Separate Account (‘‘CAC

Account’’), Valley Forge Life Insurance

Company Variable Life Separate

Account (‘‘VFL Account’’), and CNA

Investor Services, Inc.

RELEVANT 1940 ACT SECTIONS: Sections

6(c), 27(a)(3), 27(c)(2), and 27(e), and

Rules 6e–3(T)(b)(13)(ii), 6e–

3(T)(b)(13)(vii), 6e–3(T)(c)(4)(v), and

27e–1 thereunder.

SUMMARY OF APPLICATION: Applicants

seek an order to the extent necessary to

permit them or any other variable life

insurance separate account established

in the future by the Companies (‘‘Future

Accounts,’’ collectively with the CAC

Account and the VFL Account, the

‘‘Accounts’’) to support certain flexible

premium variable life insurance policies

offered currently or in the future

through the Accounts (collectively,

‘‘Policies’’) to: (1) deduct from premium

payments received under the Policies a

charge that is reasonable in relation to

each Company’s increased federal tax

burden related to the receipt of such

premium payments that results from the

application of Section 848 of the

Internal Revenue Code of 1986, as

amended, (‘‘Code’’); (2) deduct sales

charges from premium payments

received in connection with Policies in

a manner that results, in some instances,

in sales charges on subsequent premium

payments exceeding sales charges on

prior premium payments; (3) compute

sales surrender charges on such

premium payments in a manner that

results, in some instances, in sales

surrender charges on subsequent

premium payments exceeding sales

surrender charges on prior premium

payments; and (4) refrain from sending

owners of Policies a written notice of

certain refund and withdrawal rights.

FILING DATE: The application was filed

on February 14, 1996.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the Secretary of

the SEC and serving Applicants with a

copy of the request, personally or by

mail. Hearing requests should be

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

August 16, 1996 and should be

accompanied by proof of service on

Applicants in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.