Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43365

proposed action, the staff considered

denial of the proposed action. Denial of

the application would result in no

change in current environmental

impacts. The environmental impacts of

the proposed action and the alternative

action are similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for the Point Beach Nuclear

Plant, Units 1 and 2.

Agencies and Persons Consulted

In accordance with its stated policy,

on July 29, 1997, the staff consulted

with the Wisconsin State official, Ms.

Sarah Jenkins of the Wisconsin Public

Service Commission, regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated January 24, 1997, as

supplemented by letter dated May 15,

1997, which are available for public

inspection at the Commission’s Public

Document Room, The Gelman Building,

2120 L Street, NW., Washington, DC,

and at the local public document room

located at The Lester Public Library,

1001 Adams Street, Two Rivers, WI

54241.

Dated at Rockville, Maryland, this 7th day

of August 1997.

For the Nuclear Regulatory Commission.

Linda L. Gundrum,

Project Manager, Project Directorate III–1,

Division of Reactor Projects—III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–21360 Filed 8–12–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice

Applications and Amendments to

Facility Operating LicensesInvolving

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 19,

1997, through August 1, 1997. The last

biweekly notice was published on July

30, 1997, (62 FR 40843).

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 September 12, 1997, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons
43366 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

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. 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.

Carolina Power & Light Company, et

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

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of amendments request: July 8,

1997

Description of amendments request:

The proposed amendments remove the

suppression chamber water volume

band from Technical Specification (TS)

3.6.2.1.a.1 while retaining the

equivalent water level band. The values

for the suppression chamber water

volume corresponding to the low and

high suppression chamber water levels

will be retained in the Bases section of

the TS and will be revised by the

proposed amendments to account for

the displacement of water due to the

planned installation of larger emergency

core cooling system suction strainers.

The revised relationship between the

high and low suppression chamber

water levels and suppression chamber

water volume will also be described in

the Updated Final Safety Analysis

Report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below: 1. The proposed amendments

do not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change revises the values of

the minimum and maximum suppression

chamber pool water volume limits. The water

inventory of the suppression chamber pool is

not a precursor of an accident and, therefore,

cannot increase the probability of an accident

previously evaluated. The pressure

suppression chamber water pool mitigates

the consequences of loss-of-coolant accidents

(LOCAs) transients [sic], and other events by

providing a heat sink for reactor primary

system energy releases. The proposed

minimum and maximum pool water volume

values will be consistent with the current

suppression chamber pool water level limits.

No changes to setpoints will be made as a

result of the proposed change. The impact of

the proposed change to the minimum and

maximum suppression chamber pool volume

limits on the suppression chamber pool

temperatures and pressures following a

design basis LOCA, an Safety/Relief Valve

(SRV) blowdown event, an Anticipated

Transient Without Scram (ATWS) event, an

Appendix R fire event, and a station blackout

event has been evaluated and does not cause

accident parameters to exceed acceptable

values. In addition, the impact the proposed

change has on the time to reach cold

shutdown when using the alternate Residual

Heat Removal (RHR) shutdown cooling

function is negligible. The potential impact

the proposed change to the suppression

chamber pool water volume limits has on

SRV line loads, SRV discharge line reflood

height, wetwell pressurization, suppression

chamber pool swell loads, vent thrust loads,

and condensation oscillation and chugging

loads was also reviewed. The change to the

suppression chamber pool water volume

limits has no significant adverse impact on

any of these parameters. As delineated above,

the capability of the suppression chamber

water pool to perform its mitigative functions

is not affected by the proposed change.

Therefore, the proposed change does not

involve a significant increase in the

consequences of an accident previously

evaluated.

2. The proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43367

The proposed change revises the values of

the minimum and maximum volume of the

suppression chamber water pool. The

proposed change will not alter any physical

mechanism by which the suppression

chamber water pool volume is maintained

between the minimum and maximum values.

The suppression chamber pool water level

will continue to be maintained between -27

and -31 inches. The suppression chamber

pool water level limits are retained in

Technical Specification (TS) 3.6.2.1.a.1, since

this is the information available to the

operators regarding the suppression chamber

pool water volume limits. These level limits

are equivalent to the suppression chamber

pool water volume limits; therefore, it is only

the presentation of the equivalency that is

being relocated to the Bases and the Updated

Final Safety Analysis Report (UFSAR). As

such, the relocated suppression chamber

pool water volume limits are not required to

be in the TS to provide adequate protection

of the public health and safety. As a result

of the proposed strainer changes, there are no

physical changes to any other suppression

chamber components or instrumentation. No

new mode of operation is introduced as a

result of the proposed change. Analyses have

been performed which conclude that the

proposed change will not affect the

operability of the equipment designed to

mitigate the consequences of an accident.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety.

The proposed change revises the values of

the minimum and maximum suppression

chamber water pool volumes. The pressure

suppression chamber water pool mitigates

the consequences of several postulated

accidents and transients by providing a heat

sink for the primary coolant system. These

accidents and events are the postulated

design basis LOCA, an SRV blowdown event,

an ATWS event, an Appendix R fire, and

station blackout events. The consequences of

the change in the suppression pool water

volume limits have been evaluated for these

events, and there is no significant reduction

in the margin of safety.

The results of the analyses for the

postulated accidents and events indicate the

temperature of the suppression chamber pool

water could increase slightly as a

consequence of the decrease in the minimum

suppression chamber pool water volume

limit. However, the suppression chamber

pool water and containment temperatures

remain within acceptable values. The impact

of the calculated increase in containment

temperature on the available Net Positive

Suction Head (NPSH) for the Residual Heat

Removal (RHR) and Core Spray pumps has

been evaluated for the postulated design

basis LOCA and indicate[s] adequate NPSH

is maintained throughout the event.

The potential impact of the proposed

change to the suppression chamber pool

water volume limits on the SRV line loads,

SRV discharge line reflood height, wetwell

pressurization, suppression chamber pool

swell loads, vent thrust loads, and

condensation oscillation and chugging loads

was evaluated with the conclusion that there

are no adverse impacts on these parameters.

In addition, a small suppression chamber

pool water temperature increase could result

due to the reduction in minimum

suppression pool volume limit in the event

reactor shutdown is conducted through a

path utilizing the suppression chamber pool.

Such a shutdown path is an alternative to the

normal RHR shutdown cooling function, and

the small potential increase in temperature

results in a negligible increase in the time

required to reach cold shutdown conditions.

Cold shutdown conditions can still be

reached well within the Technical

Specification requirements.

The proposed increase in the suppression

pool water volume limit does not adversely

impact containment parameters as a result of

postulated accidents and events. The

potential increase in temperature of the

pressure suppression chamber pool water

does not significantly decrease the ability to

maintain containment parameters within

acceptable limits. The potential increase in

time to reach cold shutdown conditions

utilizing the suppression pool as an

alternative to the normal RHR shutdown

cooling function is negligible. Therefore, the

proposed change to revise the minimum and

maximum suppression water pool volumes

does not involve a significant reduction in a

margin of safety.

The suppression chamber pool water level

limits are retained in TS 3.6.2.1.a.1, since

this is the information available to the

operators regarding the suppression chamber

pool water volume limits. These level limits

are equivalent to the suppression chamber

pool water volume limits and the

equivalency is being relocated to the Bases

and the UFSAR. As such, the relocated

suppression chamber pool water volume

limits are not required to be in the TS to

provide adequate protection of the public

health and safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Gordon E.

Edison, Acting

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: July 21,

1997

Description of amendment request:

Technical Specification Change Request

Concerning Emergency Feedwater

Surveillance Testing. This request is to

make several changes to the ANO-2

Technical Specifications including an

extension of the emergency feedwater

(EFW) pump surveillance testing

frequency, a reduction in the minimum

steam generator pressure required to

perform the surveillance testing on the

turbine-driven EFW pump, and a

modification to the EFW pump testing

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 not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

The proposed changes included in this

amendment request are being made to the

emergency feedwater (EFW) system technical

specification (TS) surveillances. These

changes include surveillance interval

modifications, allowances to perform the

turbine driven EFW pump surveillance at a

lower steam generator (S/G) pressure,

removing the requirements to perform

specific EFW surveillance requirements (SRs)

during plant shutdowns, bases changes, and

various administrative changes. These

changes are consistent with the applicable

SRs located in NUREG-1432 and have

therefore, been previously approved by the

NRC.

These changes do not alter the functional

characteristics of any plant component and

do not allow any new modes of operation of

any component. The accident mitigation

features of the plant are not affected by the

proposed amendment request. No

modifications have been made to the EFW

system due to this amendment request.

Although the minimum steam generator

pressure has been reduced for the turbine

driven EFW pump testing, calculations show

that significant margin exists between the

proposed value and that needed to

adequately perform the test. The capability of

the EFW pumps to perform their required

safety function is not impacted by this

change. The addition of the electric driven

EFW flow path verification will help [to]

assure proper alignment of both trains of

EFW following extended outages.

The accident mitigation features of the

plant are not affected by the proposed

amendment. No modification has been made

to the pump or turbine driver. The capability

of the turbine driven EFW pump to perform

its required function is not impacted by this

change. The EFW pumps will be tested in

accordance with the more restrictive of the
43368 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

data points required by the safety analysis or

the inservice testing program. Therefore, this

change does not involve a significant

increase in the probability or consequences

of any accident previously evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

No new possibility for an accident is

introduced by modifying the proposed

specifications for the surveillance testing of

the EFW pumps. The EFW surveillance

requirements will continue to demonstrated

the pump’s ability to perform its safety

function. The modifications to the proposed

EFW surveillance requirements are

consistent with the current revision of NRC

approved NUREG -1432, ‘‘Standard

Technical Specifications Combustion

Engineering Plants’’ (ITS). Therefore, this

change does not create the possibility of a

new or different kind of accident from any

previously evaluated.

3. Does Not Involve a Significant

Reduction in Margin of Safety.

The safety function of the EFW system is

not altered as a result of this change. The

capability of the EFW pumps to perform their

required function is not impacted by this

change. The capability of the EFW pumps is

not impacted by this change. The EFW

pumps will be tested and proven operable in

accordance with the more restrictive of the

data points required by the safety analysis of

the inservice testing program. The addition of

the electric driven EFW flow path

verification will help assure [to] proper

alignment of both trains of EFW following

extended outages. Therefore, this change

does not involve a significant reduction in

the margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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: James Clifford,

Acting

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: June 26,

1997

Description of amendment request:

The proposed amendment would revise

the Operating License No. DPR-72,

License Condition 2.C.(5) and delete the

requirement for installation and testing

of flow indicators in the emergency core

cooling system to provide indication of

40 gallons per minute flow for boron

dilution from the license. Approval of

this amendment will allow removal of

the appropriate flow indicators, DH-45-

Fl and DH-46-Fl, from the Crystal River

3 (CR3) Final Safety Analysis Report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

The change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

This license amendment removes the

requirement for flow indication on the DH

drop line and auxiliary pressurizer spray line

for boron precipitation mitigation during a

LOCA [Loss of Coolant Accident]. The

original need for these indicators was to

provide flow indication to the operator to aid

in decision making relative to an alternate

active method for boron precipitation

prevention. Alternate active methods have

been replaced by the passive flow path

through the gaps which exist between the

reactor vessel and the reactor vessel

internals. Since auxiliary pressurizer spray

flow is no longer used, and no other active

means is required to be employed by the

operator in the event drop line flow is not

indicated, the original usefulness of and need

for this indication no longer exists. Removal

of this requirement from the license

condition does not involve a change in the

Improved Technical Specifications. The

operators do not use the flow indication for

decision making in post-accident conditions.

Since these instruments are no longer used

for boron precipitation mitigation during a

LOCA, abandonment or removal of flow

indicator DH-45-Fl and DH-46-Fl does not

increase the probability of an accident

because no previously evaluated accidents at

CR-3 are initiated by DH-45-Fl or DH-46-Fl.

Those CR-3 accidents that are analyzed are

contained in the Final Safety Analysis Report

(FSAR) and include events such as Loss-ofCoolant Accidents, Main Steam Line Breaks,

Station Blackout, Anticipated Transients

Without Scram, etc. Since DH-45-Fl and DH46-Fl are attached to the outside of the DH

drop line and auxiliary pressurizer spray

line, their removal will not change the

design, material, or construction standards

applicable to the DH System piping. The

removal of the indicator will not affect

overall system performance of the ECCS. All

of these previously evaluated accidents

described in the CR-3 FSAR have dose

consequences which remain well within the

requirements of 10 CFR Part 100 (25 rem

whole body, 300 rem thyroid) and GDC

[General Design Criterion] 19 (5 rem whole

body, or its equivalent to any part of the

body). Removal of DH-45-Fl and DH-46-Fl

will not alter any assumptions made in

evaluating the radiological consequences of

any accident described in the FSAR nor will

it affect any fission product barriers since the

ECCS and containment systems will still

perform to meet design requirements.

Therefore, removal of DH-45-Fl and DH-46-

Fl will not alter the consequences of an

accident previously evaluated.

Criterion 2

The change does not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

The proposed license amendment removes

the requirement for indicators which were

originally installed to aid the operator in

decision making relative to an alternate flow

path for boron precipitation mitigation

during a LOCA. These indicators no longer

serve this purpose, since alternate active flow

paths are no longer considered. Evaluations

which consider boron precipitation no longer

rely on three active methods of mitigation,

but rather one active and one passive.

Operator action is not required to effect the

backup method in the event that the primary

method fails due to a single active failure.

The flow indicators are external to the DH

System piping. They do not penetrate any

piping so their removal cannot create the

possibility of a new or different kind of

accident. The accident mitigation strategies

remain the same regardless of whether or not

the flow indicators are present. Therefore, the

flow indicators serve no purpose in the

analyses. The proposed amendment does not

affect any of the parameters or conditions

that could contribute to the initiation of any

accidents.

Criterion 3

The change does not involve a significant

reduction in the margin of safety.

Boron precipitation within the reactor

vessel during post-LOCA conditions, if it

were to occur, would challenge the margin of

safety that is provided by assuring

compliance with Criterion 5 of 10 CFR 50.46.

The license amendment does not change the

methodology of mitigating the consequences

of boron precipitation following a LOCA as

described in the current licensing basis. The

primary method of flow through the DH drop

line and the use of gap flow as the ‘‘backup’’

method for prevention of boron precipitation

have been analyzed, shown to meet all the

criteria of 10 CFR 50.46, and accepted by the

NRC. The passive method requires no

specific operator action for initiation, in the

event that the primary method fails due to a

single active failure. Therefore, the indication

serves no safety function and does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43369

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC - A5A, P. O. Box

14042, St. Petersburg, Florida 33733-

4042

NRC Project Director: Frederick J.

Hebdon

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: July 18,

1997

Description of amendment request:

The proposed amendment would revise

the Crystal River 3 (CR-3) technical

specifications (TS) to incorporate a new

TS 3.4.11 for a Low Temperature

Overpressure Protection (LTOP) System.

The proposed changes would be

consistent with the recommendations in

the NRC Generic Letter 88-11, ‘‘NRC

Position on Radiation Embrittlement of

Reactor Vessel Materials and Its Impact

on Plant Operations.’’ TS 3.5.3 and

associated TS Bases would also be

revised to reflect the proposed change.

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 Not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

This change does not involve a significant

increase in the probability or consequences

of any accident previously evaluated.

There are currently no LTOP requirements

in the CR-3 Improved Technical

Specifications. CR-3 currently implements

LTOP features through administrative

controls and a lowered PORV [poweroperated relief valve] setpoint. The proposed

change will establish new LTOP technical

specification requirements necessary to

preclude an LTOP event from occurring. The

proposed LTOP requirements are based on

safety analyses that apply ASME [American

Society of Mechanical Engineers] Code Case

N-514. These requirements will decrease the

probability of a low temperature overpressure

event by providing protection for all pressure

and temperature combinations for which a

low temperature overpressure event may be

postulated.

The consequences of a low temperature

overpressure accident are not affected by this

change. There is no change to the 10 CFR

[Code of Federal Regulations] Part 100 dose

calculation for a low temperature

overpressure accident.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated

This change does not create the possibility

of a new or different kind of accident from

any previously evaluated.

The new LTOP Technical Specification

does not require modification to the plant nor

does it create a new mode of plant operation.

The LTOP system adds no new accident

initiators.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed change does not involve a

significant reduction in the margin of safety

and will provide added safety benefit gained

through the requirements to preclude a low

temperature overpressurization event to the

RCS [reactor coolant system].

The margin of safety prior to having an

LTOP system was limited due to the

informal, administrative method of

minimizing the impact of a low temperature

overpressure accident. By formalizing these

requirements into a technical specification, at

the least, margin of safety is retained and

perhaps improved due to the elevated

significance of required actions.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC - A5A, P. O. Box

14042, St. Petersburg, Florida 33733-

4042

NRC Project Director: Frederick J.

Hebdon

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: July 29,

1997

Description of amendment request:

The proposed amendment would revise

the Crystal River Nuclear Generating

Unit 3 (CR3) technical specifications

(TS) to add subcooling margin and

decay heat removal (low pressure

injection) flow and correct certain

nomenclature in the post-accident

monitoring (PAM) instrumentation TS.

In addition, the licensee proposes to add

emergency diesel generator (EDG)

kilowatt (kW) indication to the PAM

instrumentation. Specifically, the

following TS would be revised:

A. Table 3.3.17-1, Function 8: The

descriptor is changed from

‘‘Containment Pressure (Narrow Range)’’

to ‘‘Containment Pressure (Expected

Post-Accident Range).’’

B. Table 3.3.17-1, Function 18: The

required channels for Core Exit

Temperature (Backup) is changed from

‘‘2 sets of 5’’ to ‘‘3 per core quadrant.’’

C. Table 3.3.17-1: A new Function 20

is added and designated as ‘‘Low

Pressure Injection Flow’’, with 2

required channels, and Condition E.

D. Table 3.3.17-1: A new Function 21

is added and designated as ‘‘Degrees of

Subcooling’’, with 2 required channels,

and Condition E.

E. Table 3.3.17-1: A new Function 22

is added and designated as ‘‘Emergency

Diesel Generator kW Indication’’, with 2

required channels, and Condition E. A

note clarifying the number of required

channels is added: ‘‘(c): one indicator

per EDG’’.

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 items A, B, C, D and E

corresponds to the specific TS changes

described above.

1. The proposed changes will not

significantly increase the probability or

consequences of an accident previously

evaluated because:

A/B. The changes in containment pressure

and core exit thermocouple nomenclature do

not reflect any physical changes to the

facility. This would have no impact on

accident probability or consequences.

C/D/E. The addition of low pressure

injection flow, degrees of

subcooling, and EDG kW indication to the

Post-Accident Monitoring Instrumentation

LCO [Limiting Condition for Operation] is

being done to comply with a commitment

made during the technical specification

improvement program to include in the

technical specifications that instrumentation

which monitors variables classified as Type

A in accordance with Regulatory Guide 1.97.

These three variables have been reclassified

as Type A. The associated instruments are

used in post-accident conditions to prompt

the operators to take certain mitigative

actions. Therefore, the probability of an

accident occurring is unaffected. As part of

the re-classification of these variables to Type

A and inclusion in technical specifications,

the associated monitoring instrumentation

will be under more strict surveillance and

control, which provides additional assurance

that the prescribed manual operator actions

will be implemented when necessary. This,

in turn, assures the previously evaluated

accident consequences remain valid.

2. The proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because:

A/B. The changes in containment pressure

and core exit

thermocouple nomenclature do not reflect

any physical changes to the facility. The

changes provide clarification for the

instruments which are required to comply

with the LCO. This would not create

possibility of a new or different kind of

accident.

C/D/E.The addition of low pressure

injection flow, degrees of subcooling, and

EDG kW indication to the Post-Accident

Monitoring Instrumentation LCO is being
43370 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

done to comply with a commitment made

during the technical specification

improvement program to include in the

technical specifications that instrumentation

which monitors variables classified as Type

A in accordance with Regulatory Guide 1.97.

These three variables have recently been

reclassified as Type A. The associated

instruments are used after an accident occurs

to prompt the operators to take certain

mitigative actions. Since the instrumentation

is used only post-accident, these changes do

not create the possibility of a new or different

kind of accident.

3. The proposed change will not involve a

significant reduction to the margin of safety

because:

A/B. The changes in containment pressure

and core exit thermocouple nomenclature

have no affect on the margin of safety. The

changes provide clarification of the technical

specifications. This reduces the potential for

confusion regarding this instrumentation.

C/D/E. The addition of low pressure

injection flow, degrees of

subcooling, and EDG kW indication to the

post-accident monitoring instrumentation

table in technical specifications results in

added controls on the OPERABILITY of this

post-accident monitoring instrumentation

and provides greater assurance that it will be

available should an accident occur.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC - A5A, P. O. Box

14042, St. Petersburg, Florida 33733-

4042

NRC Project Director: Frederick J.

Hebdon

Northeast Nuclear Energy Company

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

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: July 18,

1997

Description of amendment request:

The proposed amendment adds a new

Technical Specification and associated

Bases to address the operability of the

steam generator atmospheric relief

bypass valves (SGARBVs).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

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

satisfied. The proposed revision does not

involve an SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The operability of the SGARBVs provides

a method to recover from a SGTR [steam

generator tube rupture] event during which

the operator is required to perform a limited

cooldown to establish adequate subcooling as

a necessary step to limit the primary to

secondary break flow into the ruptured steam

generator. For other design events, the

SGARBVs provide a safety grade method for

cooling the unit to residual heat removal

entry conditions should the preferred heat

sink via the steam bypass system or the steam

generator atmospheric relief valves be

unavailable. This proposed revision to the

Technical Specifications will add a new

Technical Specification 3/4.7.1.6 and its

associated Bases Section 3/4.7.1.6 which

were developed bases on the information

contained in the Westinghouse Improved

Standard Technical Specifications, NUREG

1431, Rev. 1. The proposed specification and

bases provide further assurance that the

SGARBVs will be available to function as

described in the accident analysis.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This proposed revision to the Technical

Specifications to add a new specification and

bases for the SGARBVs does not cause a

change in the operation of any system or

component during normal or accident

conditions.

Therefore, the proposed revision does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed new Technical Specification

3/4.7.1.6 and its associated Bases Section 3/

4.7.1.6 were developed based on the

information contained in the Westinghouse

Improved Standard Technical Specifications,

NUREG 1431, Rev. 1. The SGARBV’s are not

currently in the Technical Specifications of

Millstone Unit No. 3 and are being added to

ensure accident mitigation functional

capability. The NUREG 1431, Rev. 1

surveillance frequency is 18 months. The

NUREG 1431, Rev. 1 surveillance frequency

bases reads ‘‘operating experience has shown

that these components usually pass the

surveillance when performed at the 18 month

frequency’’. The proposed frequency

acceptability has been evaluated by

reviewing SGARBV AWO’s [automated work

order’s] for the period from Jan. 1990 to April

1997 to confirm the absence of excessive

work orders which indicate valve functional

failures and none were identified.

Additionally, each SGARBV line consists of

one SGARBV and an associated block valve.

These proposed changes are consistent with

the design and operation of the SGARBVs.

There is no negative affect on the dose

consequences from any design basis event or

core damage frequency.

Therefore, the proposed revision does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

revision does not involve an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Deputy Director: Phillip F.

McKee

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests:

November 27, 1996

Description of amendment requests:

The proposed amendment[s] would

incorporate new steam generator tube

sleeve designs and installation and

examination techniques into the Prairie

Island Technical Specifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment[s] will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The supporting technical evaluation and

safety evaluation for the Combustion

Engineering leak tight sleeves demonstrate

that the sleeve configuration will provide

steam generator tube structural and leakage

integrity under normal operating and

accident conditions. The sleeve

configurations have been designed and

analyzed in accordance with the

requirements of the ASME [American Society

of Mechanical Engineers] Code. Mechanical

testing has shown that the sleeve and sleeve

joints provide margin above acceptance
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43371

limits. Ultrasonic examination is used to

verify the leak tightness of the above the [sic]

tubesheet sleeve welds. Testing has

demonstrated the leak tightness of the hard

roll joint as well as the structural integrity of

the hard roll joint. Tube rupture can not

occur at the hard roll joint due to the

reinforcing effect of the tubesheet. Tests have

demonstrated that tube collapse will not

occur due to postulated LOCA [loss-ofcoolant accident] loadings.

The existing Technical Specification

leakage rate requirements and accident

analysis assumptions remain unchanged in

the event that significant leakage did occur

from the sleeve joints or that a sleeve

assembly ruptured. Any leakage through the

sleeve assembly is fully bounded by the

existing steam generator tube rupture

analysis included in the Prairie Island Plant

USAR [updated safety analysis report]. The

proposed sleeving repair does not adversely

impact any other previously evaluated design

basis accident.

The sleeve minimum acceptable wall

thickness used for developing the depth

based plugging limit for the sleeve is

determined using the guidance of draft

Regulatory Guide 1.121 [≥Bases for Plugging

Degraded PWR [Pressurized-Water Reactor]

Steam Generator Tubes≥] and the pressure

stress equation of Section III of the ASME

Code. Evaluation of the minimum acceptable

wall thickness for normal, upset, and

postulated accident condition loading per the

ASME Code finds that the limiting condition

is established from normal operating

conditions which then bounds the upset and

accident condition values. Allowance for

non-destructive examination and growth of

existing sleeve wall degradation must be

made when determining the sleeve plugging

limit. The proposed plugging limit is 40%

through wall degradation. The sleeve

assembly will be examined by state of the art

non-destructive examination techniques on a

periodic basis to provide early indication of

sleeve degradation. The corrosion resistance

of the Alloy 690 sleeve has been verified by

field experience at Prairie Island. The oldest

Alloy 690 sleeves were installed May 1987.

No indication of corrosion of the sleeve or

the parent tube in the weld joint has been

identified by state-of-the-art eddy current

examination. These oldest sleeve welds did

not receive post weld heat treatment. In

addition, 5 sleeves were removed for

destructive examination in February, 1996.

No corrosion was found in any of these

sleeves including those dating from October

1992. The pulled sleeves had received post

weld heat treatment. Post weld heat

treatment can be optionally applied to the

free span sleeve weld joints to reduce the

susceptibility of the weld joint and parent

tube to stress corrosion cracking. Since the

sleeve design meets the requirements of the

ASME code and mechanical tests have

demonstrated margins above acceptance

criteria, the installation of the Combustion

Engineering leak tight sleeves will not

increase the probability or consequences of

an accident previously evaluated.

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

Installation of sleeves does not introduce

any significant changes to the plant design

basis. The use of a sleeve to span a degraded

region of steam generator tubing restores the

structural and leakage integrity of the tubing

to meet the original design bases. Stress and

fatigue analysis of the sleeve assembly shows

that the requirements for ASME Code are

met. Mechanical testing has demonstrated

that margin exists above the design criteria.

Any hypothetical accident as a result of any

degradation in the sleeved tube would be

bounded by the existing tube rupture

accident analysis.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety.

The use of the sleeves to repair degraded

steam generator tubing has been

demonstrated to maintain the integrity of the

tube bundle commensurate with the

requirements of the ASME Code and draft

Regulatory Guide 1.121 and to maintain the

primary to secondary pressure boundary

under normal and postulated accident

conditions. The safety factors used in the

verification of the strength of the sleeve

assembly are consistent with the safety

factors in the ASME Boiler and Pressure

Vessel Code used in steam generator design.

The operational and faulted condition

stresses and cumulative fatigue usage are

bounded by the ASME Code requirements.

The sleeve assembly has been verified by

testing to prevent both tube pullout and

significant leakage during normal and

postulated accident conditions. A test

program was conducted to ensure the rolled

joint design for the lower joint in the

tubesheet sleeve was leak tight and capable

of withstanding the designs loads. The

primary coolant pressure boundary of the

sleeve assembly will be periodically

inspected by non-destructive examination to

identify sleeve degradation due to operation.

Installation of sleeves will decrease the

number of tubes which must be taken out of

service. There is a small amount of primary

coolant flow reduction due to sleeves for

which an equivalent plugging sleeve to plug

ratio is assigned and is used to assess the

final equivalent plugging percentage used as

an input to other safety analyses. Because the

sleeve maintains the design basis

requirements for the steam generator tubing,

it is concluded that the proposed change

does not result in a significant reduction in

margin with respect to plant safety as defined

in the USAR or the Technical Specification

Bases.

Based on the evaluation described above,

and pursuant to 10 CFR Part 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 Part 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: John N. Hannon

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: May 15,

1997

Description of amendment requests:

The proposed amendments would

change the Technical Specifications

(TS) to revise certain limitations on

reactor coolant system leakage and

steam generator tube surveillance. The

proposed changes would implement a

voltage-based repair criteria per the

requirements of NRC Generic Letter 95-

05, ‘‘Voltage-Based Repair Criteria for

Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress

Corrosion Cracking.’’ In addition, a

typographical error in TS Section 4.12.c.

is being corrected.

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[s] will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The supporting technical evaluation and

safety evaluation for the voltage based repair

criteria demonstrate that steam generator

tube structural and leakage integrity under

normal operating and accident conditions

will be maintained. Tube burst criteria are

inherently satisfied during normal operating

conditions due to the proximity of the tube

support plate (TSP). Test data referenced in

Generic Letter 95-05 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 [≥Bases for Plugging

Degraded PWR [Pressurized-Water Reactor]

Steam Generator Tubes≥] criterion requiring

maintenance of a safety factor of 1.43 times

the MSLB pressure differential on tube burst
43372 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

is satisfied by 7/8’’ diameter tubing with

bobbin coil indications with signal

amplitudes less than the current 8.7 volts

structural limit, regardless of the indicated

depth measurement.

The upper voltage repair limit (VURL) will

be determined prior to each outage using the

most recently NRC approved 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 5.2 volts.

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

conditions to the plugging criteria. In support

of [the] 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-to

secondary leakage would result in postulated

off site and control room doses exceeding the

limits established for application of the

voltage-based repair criteria at Prairie Island.

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 [≥Standard Review Plan≥]. The

projected MSLB leak rate calculation

methodology prescribed in Section 2.b of

Generic Letter 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 and

indication growth allowance. The log-logistic

probability of leakage correlation will be

used to establish the MSLB leak rate 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[s] [do] not result in

any increase in the probability or

consequences of an accident previously

evaluated in the Updated Safety Analysis

Report (USAR).

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

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 since tubes with outside diameter

stress corrosion cracking (ODSCC) not

occurring inside the thickness of the tube

support plates will be plugged or repaired.

Neither a single or multiple tube rupture

event would be expected during all plant

conditions in a steam generator in which the

voltage based repair limit has been applied.

Northern States Power will implement a

maximum primary-to-secondary leak rate

limit of 150 gpd [gallons per day] per steam

generator to help preclude the potential for

excessive leakage during all plant conditions.

The Regulatory Guide 1.121 criterion for

establishing operational leak rate limits that

require plant shutdown are based upon leakbefore-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.

The operational leakage limit will be

reduced to 150 gpd limit consistent with

Generic Letter 95-05. This limit is expected

to provide for plant shutdown prior to

reaching critical lengths for MSLB conditions

using the lower 95% leak rate data.

Additionally, this leak-before-break

evaluation assumes that the entire crevice

area is uncovered during blowdown. Partial

uncover will provide benefit to the burst

capacity of the intersection and 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-to

secondary leakage monitoring, the possibility

of a new or different kind of accident from

any accident previously evaluated is not

created.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety.

The use of the voltage-based repair criteria

at Prairie Island maintains steam generator

tube integrity commensurate with the criteria

of the ASME [American Society of

Mechanical Engineers] Code and Regulatory

Guide 1.121. Regulatory Guide 1.121

describes a method acceptable to the

Commission for meeting GDCs [General

Design Criteria] 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 the 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 in order to assure

that radiological consequences meet the

requirements of Generic Letter 95-05.

Previous evaluations have indicated a

potential for tube deformation and collapse

during a postulated loss-of-coolant accident

(LOCA) plus safe-shutdown-earthquake (SSE)

event. The tube collapse potential arises from

TSP deformation at the support plate wedges.

Evaluation of the Westinghouse umbrella

seismic spectra provided in Westinghouse

letter NSP-92-152 for Model 51 steam

generators shows that Prairie Island is

bounded by those spectra and that no tubes

will undergo deformation due to the

combined effects of LOCA plus SSE.

Therefore, no tubes need to be excluded from

application of the voltage based criteria due

to deformation resulting from combined

LOCA plus SSE loadings. Addressing

Regulatory Guide 1.83 [≥Inservice Inspection

of Pressurized Water Reactor Steam

Generator Tubes≥] considerations,

implementation of the voltage-based repair

criteria is supplemented by enhanced eddy

current inspection guidelines to provide

consistency in voltage normalization, by an

extensive bobbin coil inspection which will

include 100% of the hot leg TSP

intersections and cold leg intersections down

to the lowest cold leg TSP with known

ODSCC, by the determination of the TSPs

having ODSCC using at least 20% random

sampling of tubes inspected over their full

length, and by rotating pancake coil

inspection (or equivalent) requirements for

the larger indications left in service 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 or

sleeves 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.

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

USAR or any Bases of the plant Technical

Specifications.

Based on the evaluation described above,

and pursuant to 10 CFR Part 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 Part 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. In addition, the proposed

correction to a typographical error has

no effect on the three standards of 10
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43373

CFR 50.92(c). 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: John N. Hannon

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company,

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

Bottom Atomic Power Station, Units

Nos. 2 and 3, York County,

Pennsylvania

Date of application for amendments:

June 4, 1997

Description of amendment request:

The proposed Technical Specifications

(TSs) amendment revises TS

Surveillance Requirement 3.8.2.1 to no

longer require that automatic emergency

diesel generator (EDG) auto-start and

trip bypass features must be functional

when the emergency core cooling

system (ECCS) is not required to be

operable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change to the facility does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change will eliminate an

inconsistency between Technical

Specifications 3.3.5.1, 3.5.2, and 3.8.2 by

clarifying that the EDG auto-start and EDG

trip bypass on ECCS initiation capability is

not required during periods in which ECCS

is not required to be OPERABLE. No physical

changes to the facility will be made per this

change. The systems, structures, and

components affected by this change are

considered to be accident mitigators and not

accident initiators. The affected systems,

structures, and components will continue to

operate within the current design parameters.

The ability of the EDGs to auto-start on a loss

of offsite power or degraded voltage will

remain unchanged. No new failure modes or

conditions adverse to safety will be created

as a result of this change. Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change to the facility does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change will eliminate an

inconsistency between Technical

Specifications 3.3.5.1, 3.5.2, and 3.8.2 by

clarifying that the EDG auto-start and EDG

trip bypass on ECCS initiation capability is

not required during periods in which ECCS

is not required to be OPERABLE. No physical

changes to the facility will be made per this

change. The systems, structures and

components affected are considered to be

accident mitigators not accident initiators.

The affected systems, structures and

components will continue to operate within

the current design parameters. No new

failure modes or conditions adverse to safety

will be created as a result of this change. The

plant conditions which do not require any

ECCS to be OPERABLE, (i.e., the plant in

MODE 5, the spent fuel storage pool gates are

removed, water level is greater than or equal

to 458 inches above reactor pressure vessel

instrument zero, and there are no OPDRVs

[operations with the potential of draining the

reactor vessel] in progress) ensure sufficient

coolant inventory to allow operator action to

prevent uncovering the fuel. The ability of

the EDGs to auto-start on a loss of offsite

power or degraded voltage will remain

unchanged. Therefore, the proposed change

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

3. The proposed change to the facility does

not involve a significant reduction in a

margin of safety.

The proposed change will eliminate an

inconsistency between Technical

Specifications 3.3.5.1, 3.5.2, and 3.8.2 by

clarifying that the EDG auto-start and EDG

trip bypass on ECCS initiation capability is

not required during periods in which ECCS

is not required to be OPERABLE. The ECCS

and EDGs capability to perform the required

safety functions as described/required in the

bases of the current plant Technical

Specifications will be maintained. Therefore,

the proposed change to the facility does not

result in a significant reduction in any

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101

NRC Project Director: John F. Stolz

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

No. 50-278, Peach Bottom Atomic

Power Station, Unit No. 3, YorkCounty,

Pennsylvania

Date of application for amendment:

June 30, 1997

Description of amendment request:

The proposed amendment would revise

the Technical Specification (TS) 2.1.1.2

safety limit minimum critical power

ratios (SLMCPRs) to be consistent with

the use of GE 13 fuel in the Unit 3 core

for operating cycle 12.

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.

The derivation of the cycle-specific

SLMCPRs for incorporation into the TS, and

its use to determine cycle-specific thermal

limits, have been performed using the

methodology discussed in ‘‘General Electric

Standard Application for Reactor Fuel,’’

NEDE-24011-P-A-13, and U.S. Supplement,

NEDE-24011-P-A-13-US, August, 1996, and

the ‘‘Proposed Amendment 25 to GE

Licensing Topical Report NEDE-24011-P-A

(GESTAR II) on Cycle Specific Safety Limit

MCPR.’’ Amendment 25 was submitted by

GENE to the U.S. Nuclear Regulatory

Commission (USNRC) on December 13, 1996.

This change in SLMCPRs cannot increase the

probability or severity of an accident.

The basis of the SLMCPR calculation is to

ensure that greater than 99.9% of all fuel rods

in the core avoid transition boiling if the

limit is not violated. The new SLMCPRs

preserve the existing margin to transition

boiling and fuel damage in the event of a

postulated accident. The fuel licensing

acceptance criteria for the SLMCPR

calculation apply to PBAPS, Unit 3, Cycle 12

in the same manner as they have applied

previously. The probability of fuel damage is

not increased. Therefore, the proposed TS

changes do not involve an increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The SLMCPR is a TS numerical value,

designed to ensure that transition boiling

does not occur in 99.9% of all fuel rods in

the core during the limiting postulated

accident. It cannot create the possibility of

any new type of accident. The new SLMCPRs

are calculated using methodology discussed

in ‘‘General Electric Standard Application for

Reactor Fuel,’’ NEDE-24011-P-A-13, and U.S.

Supplement, NEDE-24011-P-A-13-US,

August, 1996, and the ‘‘Proposed
43374 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

Amendment 25 to GE Licensing Topical

Report NEDE-24011-P-A (GESTAR II) on

Cycle Specific Safety Limit MCPR.’’

Amendment 25 was submitted by GENE to

the U.S. Nuclear Regulatory Commission

(USNRC) on December 13, 1996.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident, from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the TS

Bases will remain the same. The new

SLMCPRs are calculated using methodology

discussed in ‘‘General Electric Standard

Application for Reactor Fuel,’’ NEDE-24011-

P-A-13, and U.S. Supplement, NEDE-24011-

P-A-13-US, August, 1996, and the ‘‘Proposed

Amendment 25 to GE Licensing Topical

Report NEDE-24011-P-A (GESTAR II) on

Cycle Specific Safety Limit MCPR.’’

Amendment 25 was submitted by GENE to

the U.S. Nuclear Regulatory Commission

(USNRC) on December 13, 1996. The fuel

licensing acceptance criteria for the

calculation of the SLMCPR apply to PBAPS

[Peach Bottom Atomic Power Station], Unit

3 Cycle 12 in the same manner as they have

applied previously. The SLMCPRs ensure

that greater than 99.9% of all fuel rods in the

core will avoid transition boiling if the limit

is not violated, thereby preserving the fuel

cladding integrity. Therefore, the proposed

TS changes do not involve a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Power Authority of the State of New

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

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: April 14,

1997

Description of amendment request:

The proposed amendment revises

Appendix A, Section 6 of the Technical

Specifications. The changes will enable

Safety Review Committee (SRC) to

review plant staff performance by

deleting the plant staff performance

requirement from Section 6.5.2.9.b and

incorporating a plant staff review

requirement in Section 6.5.2.8. The

amendment also replaces the position

title of Vice President (VP) Regulatory

Affairs and Special Projects (RASP) with

Director of Regulatory Affairs and

Special Projects.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously analyzed?

Response:

This amendment application does not

involve a significant increase in the

probability or consequences of an accident

previously analyzed. The proposed changes

allow the SRC to perform a review, rather

than an audit, of plant staff performance.

This change does not diminish the SRC’s

effectiveness. A review of the 1995 QA

[quality assurance] audit of plant staff

performance shows that no findings related

to plant staff performance were issued. This

indicates that the other review mechanisms

currently in place are sufficient to ensure that

plant staff performance is monitored.

The position title change of VP-RASP to

Director-RASP is an administrative change as

all previously performed functions are being

maintained. Therefore, the proposed changes

do not affect the probability or consequences

of any previously analyzed accident.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response:

This amendment application does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed changes

affect an SRC audit requirement and a

management position title. These changes do

not affect plant equipment or the way the

plant operates. Therefore, they cannot create

a new or different kind of accident.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

This amendment application does not

involve a significant reduction in a margin of

safety. The requested Technical Specification

revisions require the SRC to review rather

than audit facility staff performance and will

not diminish the effectiveness of the SRC. A

review of the 1995 audit confirms that

performance of the annual audit is redundant

as no findings or recommendations

concerning plant staff performance were

made. The QA/ORG quarterly trend reports

and SRC review of facility staff performance

are adequate to ensure that plant staff

performance is properly monitored.

The position title change (VP-RASP to

Director-RASP) is an administrative change

as all previously performed functions are

being maintained. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposed to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019

NRC Project Director: Alexander W.

Dromerick, Acting Project Director

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: March

31, 1997, as supplemented by letter

dated July 16, 1997. The July 16, 1997,

supplement supersedes the March 31,

1997 application.

Description of amendment request:

The proposed amendment would

provide changes to Technical

Specification (TS) 2.1.2, ‘‘THERMAL

POWER, High Pressure and High Flow,’’

ACTION a.1.c for TS 3.4.1.1,

‘‘Recirculation Loops,’’ and the Bases

for TS 2.1, ‘‘Safety Limits.’’ These

changes are being made to implement

an appropriately conservative Safety

Limit Minimum Critical Power Ratio, to

include Cycle 8 specific analyses, for all

Hope Creek core and fuel designs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The derivation of the revised SLMCPRs for

Hope Creek for incorporation into the

Technical Specifications, and its use to

determine cycle—specific thermal limits,

have been performed using NRC approved

methods. Additionally, interim

implementing procedures which incorporate

cycle—specific parameters have been used

which result in a more restrictive value for

SLMCPR. These calculations do not change

the method of operating the plant and have

no effect on the probability of an accident

initiating event or transient.

There are no significant increases in the

consequences of an accident previously

evaluated. The basis of the MCPR Safety

Limit is to ensure that no mechanistic fuel

damage is calculated to occur if the limit is

not violated. The new SLMCPRs preserve the
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43375

existing margin to transition boiling and the

probability of fuel damage is not increased.

Therefore, the 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.

The proposed changes contained in this

submittal result from an analysis of the Cycle

7 and Cycle 8 core reloads using the same

fuel types as previous cycles. These changes

do not involve any new method for operating

the facility and do not involve any facility

modifications. No new initiating events or

transients result from these changes.

Therefore, the proposed Technical

Specification changes do not create the

possibility of a new or different kind of

accident, from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the

Technical Specification bases will remain the

same. The new SLMCPRs are calculated

using NRC approved methods which are in

accordance with the current fuel design and

licensing criteria. Additionally, interim

implementing procedures, which incorporate

cycle—specific parameters, have been used.

The MCPR Safety Limit remains high enough

to ensure that greater than 99.9% of all fuel

rods in the core will avoid transition boiling

if the limit is not violated, thereby preserving

the fuel cladding integrity. Therefore, the

proposed Technical Specification changes do

not involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Attorney for licensee: J. J. Keenan,

Esquire, Nuclear Business Unit - N21,

P.O. Box 236, Hancocks Bridge, NJ

08038

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: April 1,

1997, as supplemented by letter dated

May 30, 1997

Description of amendment request:

The proposed amendment would

provide changes to Technical

Specifications (TSs) 4.6.1.1, ‘‘Primary

Containment Integrity,’’ 3/4.6.1.2,

‘‘Primary Containment Leakage,’’ 3/

4.6.1.3, ‘‘Primary Containment Air

Locks,’’ 4.6.1.5.1, ‘‘Primary Containment

Structural Integrity,’’ and 4.6.1.8.2,

‘‘Drywell and Suppression Chamber

Purge System.’’ The amendment would

also change the Bases for 3/4.6.1.2,

‘‘Primary Containment Leakage,’’ 3/

4.6.1.3, ‘‘Primary Containment Air

Locks,’’ 3.4.6.1.5, ‘‘Primary Containment

Structural Integrity,’’ Section 6,

‘‘Administrative Controls,’’ and License

Condition 2.D of Facility Operating

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

‘‘Primary Containment Leakage Rate

Testing Program,’’ would be added.

These changes modify the TSs and the

Facility Operating License to adopt the

performance based containment leak

rate testing requirements (Option B) of

10 CFR Part 50, Appendix J.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Containment leak rate testing is not an

initiator of any accident. The proposed

changes do not make any physical changes to

the containment and do not affect reactor

operations or the accident analyses.

Therefore, the proposed changes do not

involve a significant increase in the

probability of any previously evaluated

accident.

Since the allowable leakage rate is not

being changed and since the analysis

documented in NUREG-1493, ‘‘PerformanceBased Containment Leak-Test Program’’

concludes that the impact on public health

and safety due to extended intervals is

negligible, the proposed changes will not

involve a significant increase in the

consequences of any previously evaluated

accident.

Therefore, adoption of a performancebased leakage testing requirements will

provide an equivalent level of safety and

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

No physical changes are being made to the

plant, nor are there any changes being made

to the operation of the plant as a result of the

proposed changes. In addition, no new

failure modes of plant equipment previously

evaluated are being introduced.

Therefore, the proposed amendment will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes are based on NRCaccepted provisions and maintain adequate

levels of reliability of containment integrity.

The performance-based approach to leakage

rate testing recognizes that historically good

results of containment testing provide

appropriate assurance of future containment

integrity. This supports the conclusion that

the impact on the health and safety of the

public as a result of extended test intervals

is negligible. Since the analysis documented

in NUREG-1493 confirms that the

performance based schedule continues to

maintain a minimal impact on public risk, it

can be concluded that the margin of safety is

not significantly affected by the proposed

changes.

Therefore, the proposed amendment will

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit - N21, P.

O. Box 236, Hancocks Bridge, New

Jersey 08038

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: July 3,

1997

Description of amendment request:

The proposed amendment would

change Technical Specification Table

3.6.3-1, ‘‘Primary Containment Isolation

Valves’’ to add valves to the list, therein.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The accidents previously evaluated in the

UFSAR [Updated Final Safety Analysis

Report] that could be possibly affected by

this proposal are those involving loss of

coolant scenarios such as a piping or

instrument line break. The proposed relief

valves, associated piping and the affected

portions of containment penetration piping

are not initiators of those accidents evaluated

in the UFSAR. The proposed relief valves

limit the post-accident maximum expected

pressures of the affected piping segments

within ASME [American Society of

Mechanical Engineers] code allowables and

system design pressures. The modification

does not cause any system or component to

be operated outside of their design rating
43376 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

allowed by applicable codes. The proposed

relief valves will be safety-related and

Seismic Category I components (except for

the relief valve discharge piping, which will

be non-safety related and seismically

analyzed, and will meet the design, material

and construction standards applicable to the

affected piping segments[)].

The proposed modifications do not

jeopardize the capability of the containment

isolation valves in the affected penetrations

to close on the receipt of a containment

isolation signal or to mitigate the

consequences of design basis accidents

evaluated in the UFSAR. Although the

modifications will result in system pressures

to be above their currently established design

values, the new peak operating pressures of

the affected piping segments will be limited

to within the requirements of the ASME

code. The modification will not alter any

assumptions previously made or change,

degrade, or prevent actions described in or

assumed in evaluating the radiological

consequences of the postulated design basis

accidents. Containment structure

temperature and pressure limits will not be

exceeded with this modification and the

offsite dose consequences will not be

affected.

Therefore these changes will not

significantly increase the probability of an

accident previously evaluated, nor involve a

significant increase in the consequences of an

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Accidents or malfunctions of equipment

important to safety previously evaluated in

the UFSAR relating to the proposed

modification involve the single active failure

of a containment isolation valve to close

upon receipt of a containment isolation

signal or its failure to limit the containment

bypass leakage following its closure. The

proposed modification: 1) does not impact

the automatic closure times of the

containment isolation valves; 2) does not

impact their capability to maintain leak

tightness during a postulated design basis

accident; and 3) does not adversely impact

the manner in which any system is operated.

The proposed modification does not

compromise the UFSAR accident analysis

assumptions and/or limits. The licensing

basis safety analysis limits for all systems

important to safety continue to be met.

Furthermore, there is no change in plant

testing proposed in this change request

which could initiate an event. Therefore,

these changes will not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed modifications and Technical

Specification changes do not change the

design limits, acceptance criteria or accident

analysis assumptions pertaining to the

containment isolation valves, their associated

piping or any other safety-related systems,

structures or components. The proposed

modification does not impact the automatic

closure times of the containment isolation

valves, nor does it impact their capability to

maintain leak tightness during a postulated

design basis accident. For the systems

affected by these penetration modifications,

there is no change in system function or

structural integrity introduced with these

proposed changes. Therefore, the changes

contained in this request do not result in a

significant reduction in a margin of safety for

the containment isolation capability of Hope

Creek.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Attorney for licensee: J. J. Keenan,

Esquire, Nuclear Business Unit - N21,

P.O. Box 236, Hancocks Bridge, NJ

08038

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: July 7,

1997

Description of amendment request:

The proposed amendment would

change Technical Specification (TS) 3/

4.8.4.2, ‘‘Motor Operated Valves -

Thermal Overload Protection

(BYPASSED),’’ to relocate the list of

applicable valves (TS Table 3.8.4.2-1) to

the Hope Creek (HC) Generating Station

Updated Final Safety Analysis Report

(UFSAR).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS revisions involve: 1) no

hardware changes; 2) no changes to the

operation of any systems or components in

normal or accident operating conditions; and

3) no changes to existing structures, systems

or components. The relocation of Technical

Specification Table 3.8.4.2-1 to the UFSAR

and existing surveillance procedures will

continue to ensure that safety-related motoroperated valves (MOVs) are capable of

performing their intended safety functions.

Therefore these changes will not significantly

increase the probability of an accident

previously evaluated. To the extent

practicable, these proposed changes were

developed consistent with the changes

approved by the NRC when developing

NUREG-1433, ‘‘Standard Technical

Specifications, General Electric Plants, BWR/

4’’, with the intent of having this relocated

information controlled in other plant

documents subject to 10CFR50.59 provisions.

Since the plant systems associated with these

proposed changes will still be capable of: 1)

meeting all applicable design basis

requirements; and 2) retain the capability to

mitigate the consequences of accidents

described in the HC UFSAR, the proposed

changes were determined to be justified.

Therefore, these changes will not involve a

significant increase in the consequences of an

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Relocation of Technical Specification

Table 3.8.4.2-1 to the UFSAR will not

adversely impact the operation of any safety

related component or equipment. Since the

proposed changes involve: 1) no hardware

changes; 2) no changes to the operation of

any systems or components; and 3) no

changes to existing structures, systems or

components, there can be no impact on the

occurrence of any accident. To the extent

practicable, these proposed changes were

developed consistent with the changes

approved by the NRC when developing

NUREG-1433, ‘‘Standard Technical

Specifications, General Electric Plants, BWR/

4’’, with the intent of having this relocated

information controlled in other plant

documents subject to 10CFR50.59 provisions.

Furthermore, there is no change in plant

testing proposed in this change request

which could initiate an event. Therefore,

these changes will not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Relocation of Technical Specification

Table 3.8.4.2-1 to the UFSAR is consistent, to

the extent practicable, with the changes

approved by the NRC when developing

NUREG-1433, ‘‘Standard Technical

Specifications, General Electric Plants, BWR/

4’’. The MOV thermal overload protection

table will reside in the UFSAR and will

ensure that the associated MOVs will be

capable of performing their intended safety

functions. Any changes to this UFSAR table

will be subject to the provisions of

10CFR50.59 and a separate safety evaluation

would be developed to support any proposed

changes that would subsequently be made.

Therefore, the changes contained in this

request do not result in a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Attorney for licensee: J. J. Keenan,

Esquire, Nuclear Business Unit - N21,
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43377

P.O. Box 236, Hancocks Bridge, NJ

08038

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260 and 50-296, Browns

Ferry Nuclear Plant, Units 1, 2 and 3,

Limestone County, Alabama

Date of amendment request: June 2,

1997 (TS 387)

Description of amendment request:

The proposed amendment allows

continued plant operation with a single

reactor recirculation loop in service.

The Nuclear Regulatory Commission

has previously determined single loop

operation is generically acceptable as set

forth in Generic Letter 86-09,

‘‘Technical Resolution of Generic Issue

B-59-(N-1) Loop Operation in BWRs

[boiling water reactors] and PWRs

[pressurized-water reactors].’’ Single

loop operation is also recognized as a

standard mode of operation in the BWR/

4 Improved Standard TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

An analysis of the limiting operational

transients has been performed by GE [General

Electric] for BFN as documented in NEDO24236 to demonstrate adequate margin to the

Safety Limit Minimum Critical Power Ratio

(SLMCPR). In addition, SLO [single loop

operation] has been specified as a operating

option for the transient and accident

evaluations performed as part of the cyclespecific core reload analyses for Units 2 and

3 which ensure that operating limit

Minimum Critical Power Ratios (OLMCPRs)

for the current fuel types are established that

maintain required margin to the fuel cladding

safety limit. A cycle-specific analysis with

SLO will be performed for Unit 1 prior to

restart and experience indicates similar

results are expected as those for Units 2 and

3.

A review of the values used in the

statistical analysis used in the basis of the

fuel cladding safety limit determined that,

due to increased uncertainties in total core

flow readings and Traversing In-Core Probe

(TIP) readings during SLO, an increase in the

SLMCPR of .02 is bounding when in SLO.

Therefore, while operating in single-loop

mode, an additional .02 is added to the

OLMCPR which maintains the same margin

to the fuel cladding safety limit as that

established for two-loop operation. This is a

conservative approach because the two-loop

transients have been shown to be more severe

than the equivalent single-loop events and,

therefore, the OLMCPRs established for twoloop operation would always be bounding.

Thus, the margin of safety for fuel clad

integrity is assured and the probability or

consequences associated with reactor

transients is not increased for SLO.

SLO results in backflow through the jet

pumps in the inactive recirculation loop

which perturbs the relationship between the

core flow and recirculation drive flow on

which the flow biased Average Power Range

Monitor (APRM) and Rod Block Monitor

(RBM) setpoint equations are based. To

compensate, the proposed TS [Technical

Specification] changes modify the setpoint

equations to correct for one-loop operation.

With this adjustment, the setpoint equations

preserve the original relationship between

the setpoints and the effective recirculation

drive flow such that the consequences of a

RWE [rod withdrawal event] in SLO are

bounded by the cycle-specific RWE analyses.

Therefore, these changes do not increase the

probability or consequences of the RWE

transient previously evaluated.

Average Planar Linear Heat Generation

Rate (APLHGR) limits are established to

ensure the acceptance criteria for fuel and

Emergency Core Cooling Systems established

in 10 CFR 50.46 are met. A SLO Loss of

Coolant Accident (LOCA) analysis was

performed using the SAFER/GESTR

computer code as documented in NEDC32484P, Revision 1, ‘‘Browns Ferry Nuclear

Plant, Units 1, 2, and 3, SAFER/GESTRLOCA, Loss-of-Coolant Accident Analysis.’’

The LOCA [loss of cooling accident] results

for SLO using SAFER/GESTR showed that,

with the application of an APLHGR

multiplier as proposed in the TS change, the

LOCA peak clad temperature for SLO will

always be lower than that for limiting design

basis pipe break for two-loop operation. An

APLHGR multiplier of 0.9 is applicable for

all current fuel types being used. This

multiplier is documented in each cyclespecific reload analysis and included in the

COLR [core operating limits report]. NEDC32484P Revision 1 also concludes that the

design basis accident (large breaks) are more

affected than small break sequences and,

therefore, the large break results are

bounding for SLO.

The Recirculation Pump Seizure event in

SLO was evaluated in NEDO-24236 and

shown to be a non-limiting event. This

conclusion is also supported by GE analyses

on other BWRs.

In summary, based on the above

discussion, the proposed changes for SLO do

not increase the probability or consequences

of an accident previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Although the proposed change allows

extended operation in a configuration that

was previously allowed for a limited period,

analysis has shown (as described in item A

above), that operation with one recirculation

pump out-of-service is within existing

analyses based on the proposed TS

requirements. Therefore, the proposed

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change to operate in singleloop recirculation mode has been analyzed in

accordance with established transient and

accident methodologies, and margins of

safety for the design basis accidents and

transients analyzed in Chapter 14 of the BFN

UFSAR [updated final safety analysis report]

have not been significantly reduced. The

basis for this conclusion is outlined in item

A above. Therefore, the proposed change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: June 9,

1997

Description of amendment request:

The amendment proposes to update the

Technical Specifications, Section 6.0, to

add a reference to NRC-approved

methodologies which will be used to

validate or generate the operating limits

in the Vermont Yankee Core Operating

Limits Report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change will not involve

any significant increase in the probability or

consequences of an accident. The change

updates the Technical Specifications to

include [an] NRC approved method reference

to allow calculation of thermal hydraulic

stability limits. It does not affect plant

operation and will not weaken or degrade the

facility.

2. The proposed change will not create the

possibility of a new or different kind of

accident since the change is administrative.

No physical alterations of the plant, setpoint

changes, or operating conditions are

proposed.

3. The proposed change will not involve a

significant reduction in a margin of safety.

The change involves an update to the

Administrative Controls in Section 6.0 of the

Technical Specifications by adding a

reference to NRC approved methods. This

administrative change does not alter plant

safety margins.
43378 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

The NRC staff has reviewed the licensee’s

analysis and, based onthis 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

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

Ropes and Gray, One International

Place, Boston, MA 02110-2624

NRC Project Director: Ronald B.

Eaton, Acting

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of amendment request: June 9,

1997

Description of amendment request:

The proposed amendments authorize a

revision to the realistic dose values for

the process gas system rupture in

Section 15.0 of the Byron/Braidwood

(B/B) Updated Final Safety Analysis

Report (UFSAR). During preparation of

a UFSAR change package, ComEd

discovered that the Final Safety

Analysis Report (FSAR) had not been

updated to correct an error from the

previous revision of the dose

calculation. Since the correct dose value

is greater than that previously reported,

the consequences of the accident had

increased, and an unreviewed safety

question resulted.

Date of publication of individual

notice in Federal Register: July 10, 1997

(62 FR 37079).

Expiration date of individual notice:

August 11, 1997 (as corrected (62 FR

39282)).

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: June 27,

1997, as supplemented by letter dated

July 2, 1997 The supplemental letter

provided clarifying information and did

not change the initial proposed no

significant hazards consideration

determination.

Brief description of amendment

request: These amendments clarify, in

the technical specifications (TSs) for

each unit, the methodology used to

satisfy surveillance requirements for the

laboratory analysis of activated carbon

(charcoal) samples from the standby gas

treatment system (SGTS) and the control

room emergency outside air supply

system (CREOASS). The specific

changes are made to Sections 4.6.5.3.b.2

and 4.6.5.3.c for the SGTS and to

Sections 4.7.b.2 and 4.7.2.c for the

CREOASS, to include a reference to

American Society for Testing Materials

(ASTM), ‘‘Radioiodine Testing of

Nuclear-Grade Gas Phase Adsorbents,’’

ASTM D3803-79. Date of publication of

individual notice in Federal Register:

July 8, 1997 (62 FR 36580)

Expiration date of individual notice:

August 7, 1997

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

April 14, 1997

Brief description of amendments: The

amendments revise Technical

Specification (TS) 3/4.3.8, ‘‘Feedwater/

Main Turbine Trip System Actuation

Instrumentation’’ by changing the

minimum channels required from three

to four. This change reflects a

modification that is being installed to

add an auxiliary contact to the trip

system logic. In addition, the

amendments revise the TS action

statement for inoperable channels to be

consistent with the Improved Standard

Technical Specifications and to account

for the additional channel.

Date of issuance: July 29, 1997

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 119 and 104

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33120).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 29, 1997. No

significant hazards consideration

comments received: No.
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43379

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Consumers Energy Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

March 27, 1997, as supplemented July 7,

1997

Brief description of amendment: The

amendment revises the Palisades Plant

license and technical specifications to

reflect the licensee’s name change from

‘‘Consumers Power Company’’ to

‘‘Consumers Energy Company.’’

Date of issuance: July 21, 1997

Effective date: July 21, 1997

Amendment No.: 176

Facility Operating License No. DPR20: Amendment revised the license and

the technical specifications.

Date of initial notice in Federal

Register: April 23, 1997 (62 FR 19828)

The July 7, 1997, letter provided

supplementary information within the

scope of the original application and did

not change the NRC staff’s initial

proposed no significant hazards

considerations determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 21, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

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:

May 27, 1997

Brief description of amendments: The

amendments delete Section 4.7.13.3.a.2

of each unit’s Technical Specifications,

regarding the minimum volume and

boron concentration of borated water

available to the Standby Makeup Pump

of the Standby Shutdown System.

Date of issuance: July 21, 1997

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 160 and 152

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

the Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33121)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 21, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

February 17, 1997, as revised May 1,

1997.

Brief description of amendment:

Changes to Technical Specification (TS)

to implement 10 CFR 50, Appendix J

Option B relating to containment

leakage tests.

Date of issuance: July 24, 1997

Effective date: July 24, 1997

Amendment No.: 156

Facility Operating License No. DPR72: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1997 (62 FR

9214), as superseded June 4, 1997 (62

FR 30632) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 24, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

April 28, 1997

Brief description of amendment:

Technical Specification (TS) 3.7.6

requires that flood protection be

provided for the service water pump

cubicles and components when the

water level exceeds a specific value. The

amendment (1) adds the closing of the

service water pump cubicle sump drain

valves to the TS, (2) revises the wording

of the action statement to be consistent

with the limiting condition for

operation, and (3) revises the associated

Bases section.

Date of issuance: July 28, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 144

Facility Operating License No. NPF49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30636)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 28, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385

Northern States Power Company,

Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of application for amendment:

January 23, 1997, as supplemented

January 28, March 4, June 19, July 2,

July 16 (2 letters), July 21, and July 25,

1997

Brief description of amendment: The

amendment documents the staff’s

review and approval of the apparent

unreviewed safety questions (USQs)

associated with (1) the updated analysis

of the design-basis accident (DBA)

containment temperature and pressure

response, and (2) the reliance on

containment pressure to compensate for

the potential deficiency in net positive

suction head (NPSH) for the emergency

core cooling system (ECCS) pumps

during a DBA with the worst case

scenario assumptions. The amendment

also authorizes the licensee to change

the Technical Specification bases and

the Updated Safety Analysis Report, to

reflect the reliance of containment

pressure to compensate for the potential

deficiency in NPSH for the ECCS pumps

following a DBA.

Date of issuance: July 25, 1997

Effective date: July 25, 1997.

Implementation shall be as specified in

Appendix C to the license.

Amendment No.: 98

Facility Operating License No. DPR22: Amendment revised the license and

the licensee’s updated safety analysis

report.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6576) The June 19, 1997, submittal,

expanded the scope of the initial

submittal dated January 23, 1997, and

therefore, another notice was issued in

Federal Register on June 24, 1997 (62

FR 34086). The July 2, July 16 (2 letters),

July 21, and July 25, 1997, submittals

provided additional clarifying

information within the scope of the

application and did not change the NRC

staff’s proposed no significant hazards

considerations determination that was

based on the June 19, 1997, submittal.

Therefore, renoticing was not

warranted. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 25, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Minneapolis Public Library,
43380 Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

June 27, 1997, as supplemented by letter

dated July 2, 1997 The supplemental

letter provided clarifying information

and did not change the initial proposed

no significant hazards consideration

determination.

Brief description of amendments:

These amendments clarify, in the

technical specifications (TSs) for each

unit, the methodology used to satisfy

surveillance requirements for the

laboratory analysis of activated carbon

(charcoal) samples from the standby gas

treatment system (SGTS) and the control

room emergency outside air supply

system (CREOASS). The specific

changes are made to Sections 4.6.5.3.b.2

and 4.6.5.3.c for the SGTS and to

Sections 4.7.b.2 and 4.7.2.c for the

CREOASS, to include a reference to

American Society for Testing Materials

(ASTM), ‘‘Radioiodine Testing of

Nuclear-Grade Gas Phase Adsorbents,’’

ASTM D3803-79.

Date of issuance: July 30, 1997

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 167 and 141

Facility Operating License Nos. NPF14 and NPF-22: The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (62 FR 36580). That

notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination by July 22,

1997. No comments have been received.

The notice also provided an opportunity

to request a hearing by August 7, 1997,

but indicated that if the Commission

makes a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendment. On July 9, 1997, the

NRC staff issued a Notice of

Enforcement Discretion in order to

delay enforcement of the current,

subject, TS requirements until the NRC

could take formal action on the July 2,

1997, application. The Commission’s

related evaluation of the amendments,

finding of exigent circumstances,

consultation with the State of

Pennsylvania, and final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

July 30, 1997.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

February 11, 1997.

Brief description of amendment: This

amendment changes the Hope Creek

Technical Specification (TS) Sections 3/

4.8.1, ‘‘A.C. Sources,’’ 6.8, ‘‘Procedures

and Programs,’’ and the Bases for

Section 3/4.8, ‘‘Electrical Power

Systems,’’ to include: 1) the relocation

of existing surveillance requirements

related to diesel fuel oil chemistry; 2)

the introduction of a new program

under TS 6.8.4.e, ‘‘Diesel Fuel Oil

Testing Program≥; 3) revisions to the TS

Bases for Section 3/4.8 to incorporate

information associated with the TS

changes; and 4) editorial changes to

implement required corrections.

Date of issuance: July 24, 1997

Effective date: As of date of issuance,

to be implemented within 60 days.

Amendment No.: 100

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 26, 1997 (62 FR 14469)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 24, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

March 3, 1997, as supplemented by

letter dated May 5, 1997

Brief description of amendment: This

amendment changes Hope Creek TSs as

follows: (1) TS 3/4.3.1, ‘‘Reactor

Protection System Instrumentation,’’ TS

3/4.3.2, ‘‘Isolation Actuation

Instrumentation,’’ and TS 3/4.3.3,

‘‘Emergency Core Cooling System

Actuation Instrumentation,’’ to include

additional information concerning

response time testing; (2) TS 4.0.5 to

reference inservice inspection and test

requirements; (3) TS 3/4.6.1, ‘‘Primary

Containment,’’ and associated Bases to

reflect a design modification; (4) TS 3/

4.7.7, ‘‘Main Turbine Bypass System,’’

to specify a new operability

requirement; and (5) the Bases for TS 3/

4.8, ‘‘Electrical Power Systems.’’

Date of issuance: July 24, 1997

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 101

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33131)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 24, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

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:

February 11, 1997, as supplemented on

May 1, June 12, and July 23, 1997

Brief description of amendments: The

amendments add a new Technical

Specification, 3/4.7.10, ‘‘Chilled Water

System - Auxiliary Building

Subsystem,’’ and an associated Bases

section to address the support function

this system provides to other necessary

safety systems.

Date of issuance: July 29, 1997

Effective date: Unit 1 to be

implemented prior to entering Mode 6

from the current unit outage; Unit 2 as

of its date of issuance, to be

implemented within 10 days of

issuance.

Amendment Nos.: 199 and 182

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 12, 1997 (62 FR 11497)

The licensee’s supplemental letters

provided additional information that

did not affect the staff’s proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

July 29, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Tennessee Valley Authority, Docket No.

50-390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendment:

October 23, 1996, as supplemented
Federal Register / Vol. 62, No. 156 / Wednesday, August 13, 1997 / Notices 43381

December 11, 1996, January 31,

February 10 and 24, March 11, April 4

and 11, May 28, June 26, and July 15,

1997.

Brief description of amendment: The

amendment changes the Watts Bar

Nuclear Plant, Unit 1, Technical

Specifications (TS) to increase the spent

fuel storage capacity from 484 fuel

assemblies to 1610 fuel assemblies and

to increase the initial enrichment of the

fuel to be stored in the spent fuel storage

racks from 3.5 weight percent (wt%) to

5.0 wt%. This modification also changes

the center-to-center spacing of stored

fuel assemblies and reflects the use of

burnup credit rack modules to be

installed peripherally along the pool

walls.

The amendment, as proposed by the

licensee, would also involve the

installation of spent fuel racks in the

spent fuel cask pit for 225 storage spaces

thus increasing the total WBN spent fuel

storage capacity to 1835 spent fuel

assemblies. The licensee proposed to

provide an impact shield that would be

placed over the fuel in the cask pit

when heavy loads are moved near or

across the cask pit area. The staff is

continuing its review of this aspect of

the licensee’s proposal. Accordingly,

this amendment authorizes the

reracking and usage of the main spent

fuel pool, as proposed for a total of 1610

spent fuel spaces. However, it does not

authorize the installation of storage

racks or storage of spent fuel in the

spent fuel cask pit. The staff’s review of

that aspect of the licensee’s application

will be addressed by further

correspondence.

Date of issuance: July 28, 1997

Effective date: July 28, 1997

Amendment No.: 6

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

Date of initial notice in Federal

Register: April 2, 1997 (62 FR 15733)

The April 4, and 11, May 28, June 26

and July 15, 1997 letters provided

clarifying informaion that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in an

environmental assessment dated April

7, 1997, and a Safety Evaluation dated

July 28, 1997. No significant hazards

consideration comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

Virginia Electric and Power Company,

et al., Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

November 9, 1987, as supplemented

March 31, 1988, June 8, 1992, and

February 4, 1997

Brief description of amendments:

These amendments reformat the

operability and surveillance

requirements for the intermediate range

channels.

Date of issuance: July 30, 1997

Effective date: July 30, 1997

Amendment Nos.: 206 and 187

Facility Operating License Nos. NPF4 and NPF-7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33136)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 30, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: February

17, 1997

Brief description of amendment: The

amendment revises the technical

specifications to move Table 3.6-1,

‘‘Containment Isolation Valves’’ to Wolf

Creek Generating Station procedures. In

addition, the technical specifications

have been modified to remove all

references to Table 3.6-1. This change is

in accordance with the guidance

provided in Generic Letter 91-08,

‘‘Removal of Component Lists from

Technical Specifications,’’ dated May 6,

1991.

Date of issuance: July 23, 1997

Effective date: July 23, 1997, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 108

Facility Operating License No. NPF42: The amendment revised the

Technical Specifications and the

Operating License.

Date of initial notice in Federal

Register: April 23, 1997 (62 FR 19838)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 23, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Dated at Rockville, Maryland, this 6th day

of August, 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

Director, Division of Reactor Projects III/IV,

Office of Nuclear Reactor Regulation.

[Doc. 97–21244 Filed 8–12–97; 8:45 am]

BILLING CODE 7590–01–F

PENSION BENEFIT GUARANTY

CORPORATION

Proposed Submission of Information

Collection for OMB Review; Comment

Request; Payment of Premiums

AGENCY: Pension Benefit Guaranty

Corporation.

ACTION: Notice of intention to request

extension of OMB approval.

SUMMARY: The Pension Benefit Guaranty

Corporation (‘‘PBGC’’) intends to

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under the Paperwork Reduction Act, of

the collection of information under its

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related instructions (OMB control

number 1212–0009; expires February

28, 1998). The collection of information

also includes a certification (on

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requirements to provide certain notices

to participants under the PBGC’s

regulation on Disclosure to Participants

(29 CFR part 4011), and surveys of plan

administrators to assess compliance

with those requirements. This notice

informs the public of the PBGC’s intent

and solicits public comment on the

collection of information.

DATES: Comments should be submitted

by October 14, 1997.

ADDRESSES: Comments may be mailed to

the Office of the General Counsel, suite

340, Pension Benefit Guaranty

Corporation, 1200 K Street, NW.,

Washington, DC 20005–4026, or

delivered to that address between 9 a.m.

and 4 p.m. on business days. Written

comments will be available for public

inspection at the PBGC’s

Communications and Public Affairs

Department, suite 240 at the same

address, between 9 a.m. and 4 p.m. on

business days.

Copies of the collection of

information may be obtained without

charge by writing to the PBGC’s

Communications and Public Affairs

Department at the address given above