68 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

1These exceptions are backfits that are necessary

in order to ensure (a) that the plants provide

adequate protection to the health and safety of the

public and are in accord with the common defense

and security, or (b) to bring the plants into

compliance with the certificates, rules or orders of

the Commission, or into conformance with written

commitments by the Corporation.

specific and/or generic regulatory staff

positions on the GDPs.

Although backfits are expected to

occur and are a part of the regulatory

process, it is important for sound and

effective regulation that backfits are

conducted in a controlled process. The

NRC staff has developed NMSS Policy

and Procedures Letter 1–53 on GDP

generic and plant-specific backfitting.

Copies of this procedure can be

obtained from the Commission Public

Document Room (PDR), 2120 L Street,

NW., Washington, DC and at the Local

Public Document Rooms (LPDRs), under

Docket No. 70–7001, at the Paducah

Public Library, 555 Washington Street,

Paducah, Kentucky 42003; and under

Docket No. 70–7002, at the Portsmouth

Public Library, 1220 Gallia Street,

Portsmounth, Ohio 45662.

Appendix 1 to NMSS Policy and

Procedures Letter 1–53 provides

guidance to the NRC staff on the proper

NRC mechanisms (e.g., rulemaking) to

use in establishing or communicating

legal requirements and NRC staff

positions to certificatees. Appendix 4

contains guidance to the NRC staff for

making backfit determinations. Once a

backfit determination has been made,

and the proposed backfit does not meet

either of the 2 exception 1 given in 10

CFR 76.76(a)(4) (i) and (ii), the NRC staff

is required by 10 CFR 76.76(a)(3) to

perform a cost/benefit analysis to

determine ‘‘that there is a substantial

increase (emphasis added) in the overall

protection of the public health and

safety or the common defense and

security to be derived from the backfit

and that the direct and indirect costs of

implementation for that plant are

justified in view of this increased

protection.’’

Appendix 3 of NMSS Policy and

Procedures Letter 1–53 contains

guidance on application of the

‘‘Substantial Increase’’ Standard. This

standard provides qualitative criteria for

NRC staff to make a safety/safeguards

‘‘net benefits’’ determination of cost/

benefits for the proposed backfit where

a quantitative approach is not feasible.

NMSS Policy and Procedures Letter

1–53 is the first backfit procedure

developed for facilities other than

nuclear power reactor facilities. In

addition, the GDPs are existing facilities

which have operated under the

Department of Energy for a number of

years. Recognizing that this procedure

may be addressing new issues, the NRC

will accept public comments which

focus on specific technical contents of

the procedure.

Opportunity for Comments

The GDP backfit implementing

procedure will be used by the NRC staff

as an interim procedure pending

completion of public review and

resolution of comments on this FR

Notice. Comments will be accepted

which focus on the specific appendices

discussed above. Comments in other

areas of the procedures will be

considered if they are directly related to

the backfit issue. Procedures such as

NMSS Policy and Procedures Letters are

used by NRC as guidance to the NRC

staff on NRC’s internal management

process.

Dated at Rockville, Maryland this 17th day

of March 1997.

For the Nuclear Regulatory Commission.

John T. Greeves,

Acting Director, Office of Nuclear Material

Safety and Safeguards.

[FR Doc. 97–7641 Filed 3–25–97; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from March 3,

1997, through March 14, 1997. The last

biweekly notice was published on

March 12, 1997.

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
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 69

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 April 25, 1997, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

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: January

15, 1997

Description of amendments request:

The proposed change would revise the

values of the minimum and maximum

suppression pool water volumes

corresponding to the upper and lower

limits of the suppression water levels

specified in TS 3.6.2.1.a.1 such that the

implementation of the administrative

controls will no longer be necessary to

ensure compliance with the 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:
70 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

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

pool water volume limits. The water

inventory of the suppression chamber 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, 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 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 pool volume limits on

the suppression pool temperatures and

pressures following a design basis LOCA, an

SRV [Safety Relief Valve] 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 RHR [Residual Heat Removal]

shutdown cooling function is negligible.

The potential impact the proposed change

to the suppression pool water volume limits

has on SRV line loads, SRV discharge line

reflood height, wetwell pressurization,

suppression pool swell loads, vent thrust

loads, and condensation oscillation and

chugging loads was also reviewed. The

proposed change to the suppression pool

water volume limits has no adverse impact

on any of these parameters.

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 amendment[s] would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

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 pool water level will

continue to be maintained between -27 and

-31 inches. As a result of the proposed

change there are no physical changes to

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

would not affect the operability of 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 amendment[s 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, Safety Relief Valve

blowdown, ATWS, Appendix R fire and

station blackout events. The consequences of

the proposed change in the suppression pool

water volume limits have been evaluated for

these events.

The results of the analyses for the

postulated accidents and events indicate the

temperature of the suppression pool water

could increase slightly as a consequence of

the decrease in the minimum suppression

pool water volume limit. However, the

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 adequate NPSH is

maintained throughout the event.

The potential impact of the proposed

change to the suppression pool water volume

limits on SRV line loads, SRV discharge line

reflood height, wetwell pressurization,

suppression 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 pool water

temperature increase could result due to the

reduction in the minimum suppression pool

volume limit in the event reactor shutdown

is conducted through a path utilizing the

suppression 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 could 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 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 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: Mark Reinhart

(Acting)

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: March

14, 1997

Description of amendment request:

The proposed change revises Technical

Specification 3/4.5.4, ‘‘Refueling Water

Storage Tank,’’ and its associated Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The non-safety, non-seismic hydrotest

pump is normally maintained separated from

the RWST [Refueling Water Storage Tank] by

a safety-related, locked closed manual

operated boundary isolation valve (1CT-22).

However, performance of Technical

Specification required surveillance test OST1506, ‘‘Reactor Coolant System Isolation

Valve Leak Test - 18 Month Interval- Mode

3,’’ requires the short term use of the

hydrotest pump during plant operating

modes. Specifically, this hydrotest pump

provides a high pressure source for leak

testing the RCS [Reactor Coolant System]

pressure isolation valves in Mode 3. The test

is performed prior to entry into Mode 2, each

refueling outage, whenever flow is

established through the pressure isolation

valves, or whenever the plant has been in

cold shutdown for greater than 72 hours.

Normally, the test is completed in less than

8 hours. Due to the piping configuration, a

break in the non-seismic portion of the

piping during these planned evolutions

could result in draining the RWST below the

minimum analyzed volume. Therefore to

mitigate the consequences of a failure in the

non-seismic piping, manual actions will be

needed to isolate the break flow, (i.e., close

valve 1CT-22), prior to reducing the water

volume in the RWST below the minimum

analyzed volume.
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 71

Based on the use of a dedicated attendant

to close valve 1CT-22, the lack of significant

accessibility concerns, and the reliability of

the valve to function, it can be concluded

that 30 minutes is ample time for a valve

attendant stationed at the valve to execute

the manual action. Since the RWST volume

margin provides up to 103 minutes to

respond to the pipe failure, it is reasonable

to assume that manual actions to isolate the

postulated pipe failure can be taken before

the RWST level decreases below the

minimum analyzed volume assumed in the

safety analysis.

Therefore, there would be no increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Based on the use of a dedicated attendant

to close valve 1CT-22, the lack of significant

accessibility concerns, and the reliability of

the valve to function, it can be concluded

that 30 minutes is ample time for a valve

attendant stationed at the valve to execute

the manual action. Since the RWST volume

margin provides up to 103 minutes to

respond to the pipe failure, it is reasonable

to assume that manual actions to isolate the

postulated pipe failure can be taken before

the RWST level decreases below the

minimum analyzed volume assumed in the

safety analysis. As a result, the capability of

the RWST to perform its safety function is

not impacted.

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 the margin

of safety.

As described in the Technical

Specification Bases, the operability of the

RWST ensures that a sufficient supply of

borated water is available for injection into

the core by the emergency core cooling

system. This borated water is used as cooling

water for the core in the event of a LOCA

[loss-of-coolant accident] and provides

negative reactivty to counteract any positive

increase in reactivity caused by reactor

coolant system (RCS) cooldown. The limits

on RWST minimum volume and boron

concentration assure that: (1) sufficient water

is available within containment to permit

recirculation cooling flow to the core, and (2)

the reactor will remain subcritical in the cold

condition following mixing of the RWST and

the RCS water volumes with all shutdown

and control rods inserted except for the most

reactive control assembly. These limits are

consistent with the assumptions of the LOCA

and steam line break analyses.

Based on the use of a dedicated attendant

to close valve 1CT-22, the lack of significant

accessibility concerns, and the reliability of

the valve to function, it can be concluded

that 30 minutes is ample time for a valve

attendant stationed at the valve to execute

the manual action. Since the RWST volume

margin provides up to 103 minutes to

respond to the pipe failure, it is reasonable

to assume that manual actions to isolate the

postulated pipe failure can be taken before

the RWST level decreases below the

minimum analyzed volume assumed in the

safety analysis. As a result, the capability of

the RWST to perform its safety function is

not impacted.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart,

Acting

Commonwealth Edison Company,

Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of amendment request: February

17, 1997

Description of amendment request:

The proposed amendment would

change the required diesel generator

load during the initial 2 hours of a

surveillance run from 2625 kW and

2750 kW to 2730 kW and 2860 kW.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because of the

following:

The proposed changes represent a

correction to the emergency diesel generator

surveillance requirement. The proposed

changes are administrative in nature and do

not significantly increase the probability or

consequences of any previously evaluated

accidents for Quad Cities Station. The

proposed amendment is consistent with the

current safety analyses and represents

sufficient requirements for the assurance and

reliability of equipment assumed to operate

in the safety analysis. As such, these changes

will not significantly increase the probability

or consequences of a previously evaluated

accident.

The associated systems related to this

proposed amendment are not assumed in any

safety analysis to initiate any accident

sequence for Quad Cities Station; therefore,

the probability of any accident previously

evaluated is not increased by the proposed

amendment.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed amendment for Quad Cities

Station’s Technical Specification is required

to ensure the diesel generator is tested in

accordance with the design basis

requirements. The proposed changes do not

create the possibility of a new or different

kind of accident previously evaluated for

Quad Cities Station. No new modes of

operation are introduced by the proposed

changes. The proposed changes are

administrative in nature and maintain at least

the present level of operability. Therefore,

the proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

The associated systems related to this

proposed amendment are not assumed in any

safety analysis to initiate any accident

sequence for Quad Cities Station; therefore,

the proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3) Involve a significant reduction in the

margin of safety because:

The proposed amendment is required to

ensure the diesel generator is tested in

accordance with the design basis

requirements. The proposed changes are

administrative in nature and do not adversely

affect existing plant safety margins or the

reliability of the equipment assumed to

operate in the safety analysis. The proposed

changes have been evaluated and found to be

acceptable for use at Quad Cities based on

system design, safety analysis requirements

and operational performance. Since the

proposed changes are administrative in

nature and maintain necessary levels of

system or component reliability, the

proposed changes do not involve a

significant reduction in the margin of safety.

The proposed amendment for Quad Cities

Station will not reduce the availability of

systems required to mitigate accident

conditions; therefore, the proposed changes

do not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request:

December 24, 1996 and January 31, 1997
72 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

Description of amendment request:

Changes to Administrative Controls

section of the Technical Specifications

needed to implement revised

management responsibilities and titles

that reflect the permanently shut down

status of plant.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

In accordance with 10 CFR 50.92, CYAPCO

[Connecticut Yankee Atomic Power

Company] and NNECO [Northeast Nuclear

Energy Company] have reviewed the attached

proposed changes and have concluded that

they do not involve a Significant Hazard

consideration (SHC). The basis of this

conclusion is that the three criterion of 10

CFR 50.92 are not compromised. The

proposed changes do not involve an SHC

because the proposed changes will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

No design basis accidents are affected by

these proposed changes. The proposed

changes are administrative in nature and are

being proposed to reflect the organizational

changes which became effective December 9,

1996.

The Haddam Neck unit changes are

replacement of the Executive Vice President,

Nuclear by the Executive Vice President and

Chief Nuclear Officer along with the

replacement of the Vice President, Haddam

Neck by the Unit Director.

No safety systems are adversely affected by

the proposed changes, and no failure modes

are associated with the changes.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

There are no changes in any way that the

plants are operated due to this administrative

change. The potential for an unanalyzed

accident is not created. There is no impact

on plant response, and no new failure modes

are introduced. The proposed administrative

and editorial changes have no impact on

safety limits or design basis accidents, and

have no potential to create a new or

unanalyzed event.

3. Involve a significant reduction in a

margin of safety.

These changes do not directly affect any

protective boundaries nor do they impact the

safety limits for the protective boundaries.

These proposed changes are administrative

and editorial in nature. Therefore there can

be no reduction in the margin of safety.

The Commission has provided guidance

concerning the application of the standards

in 10 CFR 50.92 by providing certain

examples (51 FR 7751, March 4, 1986) of

amendments that are considered not likely to

involve an SHC. The changes proposed

herein are enveloped by example (1), since

they are purely administrative changes to the

technical specifications to reflect

organizational title changes and to achieve

consistence throughout the technical

specifications at Haddam Neck.

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: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270

NRC Project Director: Seymour H.

Weiss

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 17,

1996

Description of amendment request:

The proposed change request modifies

Waterford Steam Electric Station, Unit

3, Technical Specifications 3/4.7.1.3,’’

CONDENSATE STORAGE POOL,’’ by

increasing the minimum required

contained water volume from 82 percent

to 91 percent indicated level. This

proposed change is required to ensure

that the minimum useable water volume

in the Condensate Storage Pool (CSP) is

maintained greater than or equal to

170,000 gallons. The new minimum

level accounts for the minimum level

required to prevent Emergency

Feedwater pump suction line vortexing

and instrument measurement

uncertainties.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: No.

Increasing the minimum required CSP

level will insure that the minimum required

170,000 gallons of water is available for

supply to the Emergency Feedwater System.

Maintaining the minimum required water

volume will not increase the probability of

any accident previously evaluated.

Additionally, it will not affect the

consequences of any accident. Maintaining at

least 170,000 gallons of water available in the

CSP will ensure that the system remains

within the bounds of the accident analysis.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different type of

accident from any accident previously

evaluated?

Response: No.

Increasing the minimum water volume of

the CSP from 82 percent to 91 percent does

not create a possibility for a new or different

kind of accident. The CSP will be operated

in the same manner as previously evaluated.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: No.

Operation in accordance with this

proposed change will ensure that the

minimum contained water volume of the CSP

will remain at least 170,000 gallons under all

conditions. This will maintain the present

margin of safety. Therefore, the proposed

change will not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

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

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February

5, 1997

Description of amendment request:

The proposed amendment will change

Waterford Steam Electric Station, Unit

3, Technical Specifications 3.1.2.7,

3.1.2.8, 3.5.1, 3.5.4, 3.9.1, and Bases 3/

4.1.2. The proposed change will

increase the minimum boron

concentration in the Safety Injection

Tanks (SITs) and the Refueling Water

Storage Pool (RWSP) to 2050 ppm to

reflect the safety analysis for fuel Cycle

9.

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:
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 73

The Safety Injection System (SIS) is

designed to provide core cooling in the

unlikely event of a loss of coolant accident

(LOCA). The cooling must be sufficient to

prevent significant alteration of core

geometry, preclude fuel melting, limit the

cladding metal-water reaction, and remove

the energy generated in the core for an

extended period of time following a LOCA.

The SIS fluid must contain the necessary

boron concentration to maintain the core

subcritical for the duration of a LOCA.

The proposed change increases the

minimum boron concentration in the SITs

and RWSP from 1720 ppm to 2050 ppm.

Thus, the SIT/RWSP will at all times contain

sufficient borated water to provide adequate

shutdown margin. Sampling of the system

and RWSP required by the Technical

Specifications assures that the required

dissolved boron concentration is present. In

addition to its emergency core cooling

function, the SIS functions to inject borated

water into the RCS to increase shutdown

margin following a rapid cooldown of the

RCS as a result of a steam line rupture.

Operation of the safety injection system is

credited in the steam line break analysis for

causing a decrease in core reactivity. The

current minimum RWSP/SIT concentration

to be injected is 1720 ppm. Thus an increase

to 2050 ppm will have no adverse affect on

this analysis.

The Mode 5 boron dilution event identifies

that with an initial boron concentration of

1240 ppm, a Keff of 0.98, RCS partially

drained, and one charging pump operational,

the minimum possible time to criticality is

greater than 90 minutes. For all other

combinations of Keff, RCS conditions, and

number of charging pumps, the time to loss

of shutdown margin is greater than 55

minutes. Thus, the proposed increase in

boron concentration will not affect the results

of the Mode 5 boron dilution event.

The change to the action statement of TS

3.9.1 assures that the more limiting reactivity

condition of a Keff less than 0.95 or a boron

concentration of 2050 ppm specified in the

COLR [Core Operating Limit Report] will be

adhered to during refueling operations.

The upper limit on boron concentration

has not changed; therefore, there will be no

affect on boric acid precipitation post-LOCA.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

The proposed change does not physically

alter the configuration of the plant and,

therefore, does not create the possibility of a

new or different kind of accident from any

previously evaluated accident.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change maintains the

minimum of 55 minutes to criticality for the

refueling mode boron dilution event analysis.

The proposed change continues to ensure

that borated water of sufficient concentration

is injected from both the SITs and the RWSP

in the event of a LOCA or MSLB [main steam

line break] and that boric acid does not

precipitate in the core during long term

cooling following a LOCA.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

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

NRC Project Director: William D.

Beckner

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: February

10, 1997

Description of amendment request:

The proposed amendment would

provide the requirements for avoidance

and protection from thermal hydraulic

instabilities as described in NRC

Generic Letter 94-02, ‘‘Long-Term

Solutions and Upgrade of Interim

Operating Recommendations for

Thermal Hydraulic Instabilities in

Boiling Water Reactors.’’

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.

The proposed amendment does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. In fact, it does not result in an

increase in the probability or consequences

of any previously evaluated accidents. The

implementation of [Boiling Water Reactor

Owners’ Group] BWROG Long-Term Stability

Solution Option I-D at [Cooper Nuclear

Station] CNS does not modify the

assumptions contained in the existing

accident analysis. The use of an exclusion

region and the operator actions required to

avoid and minimize operation inside the

region do not increase the possibility of an

accident.

Conditions of operation outside of the

exclusion region are within the analytical

envelope of the existing safety analysis. The

operator action requirement to exit the

exclusion region upon entry minimizes the

possibility of an oscillation occurring. The

actions to drive control rods and/or to

increase recirculation flow to exit the region

are maneuvers within the envelope of normal

plant evolutions. The flow-biased scram has

been analyzed and will provide automatic

fuel protection in the event of an instability.

Thus, each proposed Technical Specification

requirement provides defense for protection

from an instability event within the existing

assumptions of the accident analysis.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

As stated above, the proposed Technical

Specification requirements either mandate

operation within the envelope of existing

plant operating conditions or force specific

operating maneuvers within those carried out

in normal operation. Since operation of the

plant with all of the proposed requirements

is within the existing operating basis, an

unanalyzed accident will not be created

through implementation of the proposed

change.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

Each of the proposed requirements for

plant thermal-hydraulic stability provides a

means for fuel protection. The combination

of avoiding possible unstable conditions and

the automatic flow-biased reactor scram

provides an in-depth means for fuel

protection. Therefore, the individual or

combination of means to avoid and suppress

an instability supplements the margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Auburn Memorial Library,

1810 Courthouse Avenue, Auburn, NE

68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

NE 68602-0499

NRC Project Director: William D.

Beckner

Northeast Nuclear Energy Company, et

al., Docket No. 50-245, Millstone

Nuclear Power Station, Unit No. 1, New

London, Connecticut

Date of amendment request: March 6,

1997

Description of amendment request:

During a self assessment, the licensee

identified weaknesses in the current

Technical Specifications regarding

allowed outage times for certain specific

protective instrumentation and also for

reactor building access control. The

proposed amendment is designed to

eliminate these weaknesses by adopting

guidance from NUREG-0123, ‘‘Standard

Technical Specifications for General

Electric Boiling Water Reactors (BWR/
74 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

5),’’ Revision 3, and NUREG-1433,

Standard Technical Specifications

General Electric Plants BWR/4,’’

Revision 1.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Millstone Nuclear

Power Station, Unit No. 1, in accordance

with the proposed amendment, will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The inherent redundancy and reliability of

the protective instrumentation trip systems

ensure that the consequences of an accident

are not significantly increased. In addition,

the restrictive Allowable Outage Time (AOT)

interval limits the probability of the

protective instrument channel being

unavailable and an accident requiring its

function from occurring simultaneously. The

requirement that the associated trip function

maintains trip capability ensures that the

protective instrumentation response will

occur such that the consequences of an

accident are not different from those

previously evaluated.

Instruments addressed in the proposed TS

respond to changes in the plant. The

proposed (AOTs) provide a two-hour interval

where the instrument is inoperable, yet the

Technical Specification (TS) Limiting

Condition for Operation (LCO) action

statement is not immediately entered. The

probability of a plant transient being initiated

by a trip of a coincident channel during

surveillance testing is reduced since the

channel under test will only be tripped for

a small portion of the test interval. Therefore,

AOTs provided by the proposed TS have no

effect on the probability of occurrence of

previously evaluated accidents.

The proposed TS changes provide a twohour interval where the instrument is

inoperable, but the TS LCO action statement

is not immediately entered. If a single failure

occurred on the other channel of the trip

system being tested and the channel being

tested was not in the trip condition, a valid

signal might not provide the required

protective action. The probability of an event

requiring initiation of the protective function

within the proposed AOT is low.

Additionally, surveillance testing is not

generally performed on multiple sensors

simultaneously. So, other trip functions and

sensors remain operable and the probability

of extensive inoperabilities affecting diverse

trip functions is low. A spurious trip of a

coincident channel could initiate a plant

transient (for example, a reactor scram or a

main steam isolation valve closure); however,

these transients are bounded by the current

analyses. Moreover, the original TS bases

submitted as part of the application for

Millstone Unit No. 1’s Provisional Operating

License (dated October 7, 1970) included

recognition that instruments would be

inoperable during required functional tests

and calibrations. Thus, these conditions were

recognized in the original design bases and

constitute part of the licensing bases of the

plant. NUREG-0123 provided specific time

frames[,] ...AOTs addressed in the table

notes[,] and specific action statements.

Millstone Unit No. 1 AOT values chosen are

consistent with these values and less than

those approved in NUREG-1433 which had a

more detailed study performed to lengthen

the AOT value.

The existing TS definition for Instrument

Functional Test would be difficult to satisfy

if the LCO condition of tripping the

inoperable channel was performed. A similar

problem of complying with the Instrument

Calibration definition also exists. The TS

requirement to perform functional tests and

calibrations is not consistent with a

requirement to trip the system under test.

The proposed TS changes permit more

complete functional and calibration testing.

For example, the main scram contactors

could be included within the surveillance

tests. Therefore, these TS clarifications do

not increase the consequences of any

previously analyzed accidents.

The two-hour instrumentation AOT for the

Air Ejector Off-Gas System radiation

monitors is slightly less restrictive than that

allowed by the NUREG-0123. Since this

requirement was relocated from NUREG1433, there is no corresponding requirement

for comparison. These radiation monitors are

arranged in a two-out-of-two logic; therefore,

both must trip to initiate the required action

(closure of the off-gas isolation valve to the

main stack). This action, however, is

automatically delayed by 15 minutes. A high

radiation condition sensed by the monitor in

service would provide sufficient time to take

corrective actions. Since a two-hour AOT is

deemed acceptable for instrumentation in

system[s] such as the Reactor Protection

System and Emergency Core Cooling

Systems, it is appropriate to apply a two-hour

AOT to these radiation monitors.

Additionally, the NUREG-0123 AOT of one

hour does not allow sufficient time to

perform required surveillance testing without

placing undue stress on the test performer.

The probability of a plant transient (e.g., loss

of condenser vacuum) resulting from a trip of

the coincident channel during surveillance

testing is reduced since the channel under

test will only be tripped for a small portion

of the test interval. This transient is bounded

by existing analyses. Therefore, this proposed

AOT will not significantly increase the

probability or consequences of an accident

previously evaluated.

Since no physical change is being made to

the secondary containment, or to any systems

or components that interface with the

secondary containment, there is no change in

the probability of any accident analyzed in

the UFSAR [Updated Final Safety Analysis

Report].

The proposed change continues to ensure

the secondary containment requirements

meet the licensing basis. Also, the proposed

changes are based on Standard Technical

Specifications, NUREG-1433, ‘‘Standard

Technical Specifications General Electric

Plants, BWR/4,’’ Revision 1 guidelines and

implement actions to be taken when

secondary containment integrity is not met.

If secondary containment integrity is not met,

existing TS 3.7.C directs the plant to be

placed in an operating condition where

secondary containment is not required, e.g.,

COLD SHUTDOWN. A four hour allowable

outage time is proposed which provides a

period of time to correct the problem that is

commensurate with the importance of

maintaining secondary containment during

RUN, STARTUP/HOT STANDBY or HOT

SHUTDOWN. The secondary containment is

not an initiator for any accident. Therefore,

the proposed change will not increase the

probability of any previously analyzed

accident. This short time period ensures that

the probability of an accident requiring

secondary containment integrity operability

occurring during periods when secondary

containment integrity is inoperable is

minimal.

The proposed surveillance requirement is

based on the NUREG-1433 surveillance

requiring periodic confirmation that at least

one door in each of the double-doored

accesses to the secondary containment is

closed, provides additional assurance of

secondary containment system integrity.

While this is a deviation from NUREG-1433

(which requires that both doors in each

access be closed except for normal entry and

exit), it is consistent with the current

definition of Secondary Containment

Integrity, which requires that at least one

door in each access opening be closed.

Hence, the deviation is justifiable and

represents increased passive testing which

will provide increased awareness of plant

conditions. Increased awareness of plant

conditions should reduce the probability or

consequences of any accident previously

evaluated.

Since the aspects of secondary

containment integrity affected by reactor

building access control are being revised in

this proposed amendment to agree with the

allowable outage time allowed by NUREG1433 upon loss of secondary containment

integrity, the change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Since the editorial items do not alter the

meaning or intent of any requirements, they

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The operation of Millstone Nuclear

Power Station, Unit No. 1, in accordance

with the proposed amendment, will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change to the protective

instrumentation trip system specifications do

not create the possibility of a new or different

kind of accident because they do not

introduce any new operational modes or

physical modifications to the plant.

Instruments addressed in the proposed TS

respond to changes in the plant. The

proposed AOTs provide a two-hour interval

where the instrument is inoperable, yet the

TS LCO action statement is not immediately

entered. Given a single failure, this could

impact the response of the trip channel but

not the initiation of the event. The only
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 75

action resulting from the AOTs is to perform

testing as required by TS. Spurious signals

during testing could initiate transients but

would be bounded by the previous transient

analyses. These tests do not subject the

instruments to any conditions beyond their

design specifications and are performed in

accordance with approved testing standards.

This testing ensures equipment operability

by identifying degraded conditions, initiating

corrective action and properly retesting them.

Therefore, the proposed TS changes will not

introduce a new or different kind of accident

than previously evaluated.

The two-hour instrumentation AOT for the

Air Ejector Off-Gas System radiation

monitors is slightly less restrictive than that

allowed by the NUREG-0123. Since this

requirement was relocated from NUREG1433, there is no corresponding requirement

for comparison. These radiation monitors are

arranged in a two-out-of-two logic; therefore,

both must trip to initiate the required action

(closure of the off-gas isolation valve to the

main stack). This action, however, is

automatically delayed by 15 minutes. A high

radiation condition sensed by the monitor in

service would provide sufficient time to take

corrective actions. Since a two-hour AOT is

deemed acceptable for instrumentation in

system[s] such as the Reactor Protection

System and Emergency Core Cooling

Systems, it is appropriate to apply a two-hour

AOT to these radiation monitors.

The proposed changes to Millstone Unit

No. 1 Technical Specifications Section 3.7/

4.7 and associated bases were developed

using the guidance provided in the Standard

Technical Specifications, NUREG-1433,

‘‘Standard Technical Specifications General

Electric Plants, BWR/4,’’ Revision 1.

Augmentation of the existing surveillance

requirements by incorporation of an

additional NUREG-1433 based surveillance,

provides additional assurance of secondary

containment system integrity. While this is a

deviation from NUREG-1433 (which requires

that both doors in each access be closed

except for normal entry and exit), it is

consistent with the current definition of

Secondary Containment Integrity which

requires that at least one door in each access

opening be closed. Hence, the deviation is

justifiable and represents increased passive

testing which will provide increased

awareness of plant conditions. Increased

awareness of plant conditions will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. Since the proposed changes do not

significantly degrade the present level of

system operability and add provisions from

NUREG-1433, the proposed amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Since the editorial items do not alter plant

configurations or operating modes, they do

not create the possibility of a new or different

kind of accident.

3. The operation of Millstone Nuclear

Power Station, Unit No. 1, in accordance

with the proposed amendment, will not

involve a significant reduction in a margin of

safety.

The protective instrumentation

surveillance requirements provide

verification of the operability of the trip

system instrumentation channels. In

addition, the channel that monitors the

identical Trip Function within the same trip

system maintains trip capability for the

relatively short duration that the coincidence

change is in effect. This ensures that

protective instrumentation reliability is

maintained. The proposed change provides

for a specific time period to perform required

surveillances on instrument channels

without trips present in associated trip

systems. This time allotment tends to

enhance the margin of safety by decreasing

the probability of unnecessary challenges to

safety systems and inadvertent plant

transients.

The proposed TS provide a two-hour

interval where the instrument is inoperable,

yet the TS LCO action statement is not

immediately entered. If a single failure

occurred on the other channel of the trip

system being tested and the channel being

tested was not in the tripped condition, a

valid signal might not provide the required

protective action. The probability of an event

requiring initiation of the protective function

within the proposed AOT is low.

Additionally, surveillance testing is not

generally performed on multiple sensors

simultaneously. So, other trip functions and

sensors remain operable and the probability

of extensive inoperabilities affecting diverse

trip functions is low.

The existing TS definition for Instrument

Functional Test would be difficult to satisfy

if the LCO condition of tripping the

inoperable channel was performed. A similar

problem of complying with the Instrument

Calibration definition also exists. Moreover,

the original TS bases submitted as part of the

application for Millstone Unit No. 1—s

Provisional Operating License (dated October

7, 1970) included recognition that

instruments would be inoperable during

required functional test and calibrations.

Thus, these conditions were recognized in

the original design bases and constitute part

of the licensing bases of the plant. NUREG0123 provided specific time frames[,]...AOTs

addressed in the table notes[,] and specific

action statements. Millstone Unit No. 1 AOT

values chosen are consistent with these

values and less than those approved in

NUREG-1433 which had a more detailed

study performed to lengthen the AOT value.

The only action resulting from the

proposed TS is to perform testing as required

by TS. Spurious signals during testing could

initiate equipment or plant transients but

would be bounded by the previous transient

analysis. These tests do not subject the

instruments to any conditions beyond their

design specifications and are performed in

accordance with approved testing standards.

This testing ensures equipment operability

by identifying degraded conditions, initiating

corrective action and properly retesting them.

Therefore, the proposed TS do not involve a

significant reduction in a margin of safety.

The two-hour instrumentation AOT for the

Air Ejector Off-Gas System radiation

monitors is slightly less restrictive than that

allowed by the NUREG-0123. Since this

requirement was relocated from NUREG1433, there is no corresponding requirement

for comparison. These radiation monitors are

arranged in a two-out-of-two logic; therefore,

both must trip to initiate the required action

(closure of the off-gas isolation valve to the

main stack). This action, however, is

automatically delayed by 15 minutes. A high

radiation condition sensed by the monitor in

service would provide sufficient time to take

corrective actions. Since a two-hour AOT is

deemed acceptable for instrumentation in

system[s] such as the Reactor Protection

System and Emergency Core Cooling

Systems, it is appropriate to apply a two-hour

AOT to these radiation monitors and does

not involve a significant reduction in the

margin of safety.

The addition of an allowable outage time

of four hours for Secondary Containment

Integrity has negligible effect on accident

occurrence or consequences. Since the

proposed change does not involve the

addition or modification of plant equipment,

is consistent with the intent of the existing

Technical Specifications, is consistent with

the current industry practices as outlined in

NUREG-1433, (except for the deviation noted

above), and is consistent with the design

basis of the plant and the accident analysis,

no action will occur that will involve a

significant reduction in a margin of safety.

Since the editorial items do not alter the

meaning or intent of any requirements, they

do not affect the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

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: July 28,

1995, as revised February 21, 1997

Description of amendment requests:

The proposed amendments would

revise the Technical Specifications

(TSs) to allow use of credit for soluble

boron in spent fuel pool criticality

analyses. The licensee’s February 21,

1997, submittal is a revision to its

original amendment requests dated July

28, 1995. The generic methodology for

crediting soluble boron in spent fuel

rack criticality analyses was approved
76 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

by the NRC on October 25, 1996.

However, because of changes made to

the generic methodology as a result of

comments from the NRC staff, it was

necessary for NSP to revise its original

amendment requests. In addition, the

licensee has revised its request by

eliminating the proposed relocation of

the spent fuel pool operating limits to

the Unit 1 core operating limits report

and will retain these limits in the TSs.

The licensee’s original application for

amendments was published in the

Federal Register on September 23, 1996,

(61 FR 49800).

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.

There is no increase in the probability of

a fuel assembly drop accident in the spent

fuel pool when considering the presence of

soluble boron in the spent fuel pool water for

criticality control. The handling of the fuel

assemblies in the spent fuel pool has always

been performed in borated water.

The criticality analysis showed the

consequences of a fuel assembly drop

accident in the spent fuel pool are not

affected when considering the presence of

soluble boron.

There is no increase in the probability of

the accidental misloading of spent fuel

assemblies into the spent fuel pool racks

when considering the presence of soluble

boron in the pool water for criticality control.

Fuel assembly placement will continue to be

controlled pursuant to approved fuel

handling procedures and will be in

accordance with the Technical Specification

spent fuel rack storage configuration

limitations. The addition of the spent fuel

pool storage configuration surveillance in

proposed Specification 4.20 will provide

increased assurance that a spent fuel pool

inventory verification will be completed in a

timely manner after completion of a fuel

handling campaign in the spent fuel pool.

There is no increase in the consequences

of the accidental misloading of spent fuel

assemblies into the spent fuel pool racks

because criticality analyses demonstrate that

the pool will remain subcritical following an

accidental misloading if the pool contains an

adequate boron concentration. The proposed

Technical Specifications limitations will

ensure that an adequate spent fuel pool boron

concentration will be maintained.

There is no increase in the probability of

the loss of normal cooling to the spent fuel

pool water when considering the presence of

soluble boron in the pool water for

subcriticality control since a high

concentration of soluble boron has always

been maintained in the spent fuel pool water.

A loss of normal cooling to the spent fuel

pool water causes an increase in the

temperature of the water passing through the

stored fuel assemblies. This causes a decrease

in water density which would result in a

decrease in reactivity when Boraflex neutron

absorber panels are present in the racks.

However, since Boraflex is not considered to

be present, and the spent fuel pool water has

a high concentration of boron, a density

decrease causes a positive reactivity addition.

However, the additional negative reactivity

provided by the proposed 1800 ppm boron

concentration limit, above that provided by

the concentration required to maintain Keff

less than or equal to 0.95 (750 ppm), will

compensate for the increased reactivity

which could result from a loss of spent fuel

pool cooling event. Because adequate soluble

boron will be maintained in the spent fuel

pool water, the consequences of a loss of

normal cooling to the spent fuel pool will not

be increased.

Therefore, based on the conclusions of the

above analysis, the proposed changes will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

Spent fuel handling accidents are not new

or different types of accidents, they have

been analyzed in Section 14.5.1 of the

Updated Safety Analysis Report.

Criticality accidents in the spent fuel pool

are not new or different types of accidents,

they have been analyzed in the Updated

Safety Analysis Report and in Criticality

Analysis reports associated with specific

licensing amendments for fuel enrichments

up to 5.0 weight percent U-235.

The Prairie Island Technical Specifications

currently contain limitations on the spent

fuel pool boron concentration. Current

Specification 3.8.E.2, which covers the

storage of restricted fuel assemblies in an

unverified condition, and Specification

3.8.B.1.c for the loading of fuel assemblies

into a cask in the spent fuel pool, contain

requirements for spent fuel pool boron

concentration. The actual boron

concentration in the spent fuel pool has

always been kept at a higher value for

refueling purposes. New Specification 3.8.E.2

establishes new boron concentration

requirements for the spent fuel pool water

consistent with the results of the new

criticality analysis (Exhibit E [of the February

21, 1997, submittal]).

Since soluble boron has always been

maintained in the spent fuel pool water, and

is currently required by Technical

Specifications under some circumstances, the

implementation of this new requirement will

have little effect on normal pool operations

and maintenance. The implementation of the

proposed new limitations on the spent fuel

pool boron concentration will only result in

increased sampling to verify boron

concentration. This increased sampling will

not create the possibility of a new or different

kind of accident.

Because soluble boron has always been

present in the spent fuel pool and is required

by current Technical Specifications as

discussed above, a dilution of the spent fuel

pool soluble boron has always been a

possibility. However, it was shown in the

spent fuel pool dilution evaluation (Exhibit

D [of the February 21, 1997, submittal]) that

a dilution of the Prairie Island spent fuel pool

which could reduce the rack Keff to less than

0.95 is not a credible event. Therefore, the

implementation of new limitations on the

spent fuel pool boron concentration will not

result in the possibility of a new kind of

accident.

Revised Specifications 3.8.E.1, 5.6.A.1.d

and 5.6.A.1.e continue to specify the

requirements for the spent fuel rack storage

configurations, the only significant changes

relate to the criteria for determining the

storage configuration. Since the proposed

spent fuel pool storage configuration

limitations will be similar to those currently

in the Prairie Island Technical Specifications,

the new limitations will not have any

significant effect on normal spent fuel pool

operations and maintenance and will not

create any possibility of a new or different

kind of accident. Verifications will continue

to be performed to ensure that the spent fuel

pool loading configuration meets specified

requirements.

As discussed above, the proposed changes

will not create the possibility of a new or

different kind of accident. There is no

significant change in plant configuration,

equipment design or equipment. The

accident analysis in the Updated Safety

Analysis Report remains bounding.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety.

The Technical Specification changes

proposed by this License Amendment

Request and the resulting spent fuel storage

operating limits will provide adequate safety

margin to ensure that the stored fuel

assembly array will always remain

subcritical. Those limits are based on a plant

specific criticality analysis (Exhibit E)

performed in accordance [with] the

Westinghouse spent fuel rack criticality

analysis methodology described in Reference

4 [in Exhibit A of the February 21, 1997,

submittal].

While the criticality analysis utilized credit

for soluble boron, a storage configuration has

been defined using a 95/95 Keff calculation to

ensure that the spent fuel rack Keff will be

less than 1.0 with no soluble boron. Soluble

boron credit is used to offset uncertainties,

tolerances and off-normal conditions and to

provide subcritical margin such that the

spent fuel pool Keff is maintained less than

or equal to 0.95.

The loss of substantial amounts of soluble

boron from the spent fuel pool which could

lead to exceeding a Keff of 0.95 has been

evaluated (Exhibit D) and shown to be not

credible.

The evaluations in Exhibit D, which show

that the dilution of the spent fuel pool boron

concentration from 1800 ppm to 750 ppm is

not credible, combined with the 95/95

calculation, which shows that the spent fuel

rack Keff will remain less than 1.0 when

flooded with unborated water, provide a

level of safety comparable to the conservative

criticality analysis methodology required by

References 1, 2 and 3 [in Exhibit A of the

February 21, 1997, submittal].
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 77

Therefore, the proposed changes in this

license amendment will not result in a

significant reduction in the plant’s margin of

safety.

Based on the evaluation above, and

pursuant to 10 CFR 50, Section 50.91,

Northern States Power Company has

determined that operation of the Prairie

Island Nuclear Generating Plant in

accordance with the proposed license

amendment request does not involve any

significant hazards considerations as defined

by NRC regulations in 10 CFR 50, Section

50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: John N. Hannon

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests:

February 14, 1997

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant (DCPP) Unit Nos. 1

and 2 to revise the surveillance

frequencies from at least once every 18

months to at least once per refueling

interval (nominally 24 months) for 8

slave relay tests, 20 electrical system

tests and 1 electrical TS Bases change,

and 5 miscellaneous tests.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS surveillance interval

increase to 24 months do not alter the intent

or method by which the inspections, tests, or

verifications are conducted; do not alter the

way any structure, system, or component

functions; and do not change the manner in

which the plant is operated.

The surveillance, maintenance, and

operating histories indicate that the

equipment will continue to perform

satisfactorily with longer surveillance

intervals. Few surveillance and maintenance

problems were identified. No problems have

recurred following identification of root

causes and implementation of corrective

actions.

There are no known mechanisms that

would significantly degrade the performance

of the evaluated equipment during normal

plant operation. All potential time related

degradation mechanisms have insignificant

effects in the timeframe of interest (24

months +25 percent, or 30 months). Based on

the past performance of the equipment, the

probability or consequences of accidents

would not be significantly affected by the

proposed surveillance interval increases.

Deletion of the phrase ‘‘during shutdown’’

for the applicable electrical TS will not alter

the intent or method by which the

inspections, tests, or verifications are

conducted; nor alter the way any structure,

system, or component functions. DCPP has

administrative programs in place which

require evaluation of risk and suitability of

surveillance and maintenance activities to

ensure that performance during plant

operation does not adversely affect safety.

The administrative change for one PORV

TS regarding channel calibration only

maintains the existing surveillance

frequency. This revision does not alter the

intent or method by which the inspections,

tests, or verifications are conducted; nor alter

the way any structure, system, or component

functions.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

For the proposed TS changes involving

surveillance interval increases to 24 months,

the surveillance and maintenance histories

indicate that the equipment will continue to

effectively perform its design function over

the longer operating cycles. Additionally, the

increased surveillance intervals do not result

in any physical modifications, affect safety

function performance or the manner in

which the plant is operated, or alter the

intent or method by which surveillance tests

are performed. No problems have recurred

following identification of root causes and

implementation of corrective actions. All

identified potential time related degradations

have insignificant effects in the timeframe of

interest. The proposed surveillance interval

increases would not affect the type of

accident possible.

Deletion of the phrase —during

shutdown— for the applicable electrical TS

does not result in any physical modifications,

affect safety function performance or the

manner in which the plant is operated, or

alter the intent or method by which

surveillance tests are performed. DCPP has

administrative programs in place which

require evaluation of risk and suitability of

surveillance and maintenance activities to

ensure that performance during plant

operation does not adversely affect safety.

The administrative change for one PORV

TS regarding channel calibration only

maintains the existing surveillance

frequency. This revision does not result in

any physical modifications, affect safety

performance or the manner in which the

plant is operated, or alter the intent or

method by which surveillance tests are

performed.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

For the proposed TS changes involving

surveillance interval increases to 24 months,

evaluation of historical surveillance and

maintenance data indicates there have been

few problems experienced with the evaluated

equipment. There are no indications that

potential problems would be cycle ength

dependent or that potential degradation

would be significant for the timeframe of

interest; therefore, increasing the surveillance

interval will have little, if any, impact on

safety. There is no safety analysis impact

since these changes will have no effect on

any safety limit, protection system setpoint,

or limiting condition for operation, and there

are no hardware changes that would impact

existing safety analysis acceptance criteria.

Safety margins would not be significantly

affected by the proposed surveillance interval

increases.

Deletion of the phrase ‘‘during shutdown’’

for the applicable electrical TS has no safety

analysis impact since these changes will have

no effect on any safety limit, protection

system setpoint, or limiting condition for

operation, and there are no hardware changes

that would impact existing safety analysis

acceptance criteria. DCPP has administrative

programs in place which require evaluation

of risk and suitability of surveillance and

maintenance activities to ensure that

performance during plant operation does not

adversely affect safety.

The administrative change for one PORV

TS regarding channel calibration only

maintains the existing surveillance

frequency. This revision has no safety

analysis impact.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120
78 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

NRC Project Director: William H.

Bateman

Portland General Electric Company, et

al., Docket No. 50-344, Trojan Nuclear

Plant, Columbia County, Oregon

Date of amendment request: January

16, 1997, as supplemented February 24,

1997.

Description of amendment request:

The proposed amendment would allow

pre-operational testing and load

handling of spent fuel transfer and

storage casks in the Trojan Fuel

Building.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

staff’s review is presented below:

The proposed changes would not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. With the permanent cessation of

operations at the Trojan Plant, the number of

potential accidents was reduced to those

types of accidents associated with the storage

of irradiated fuel and radioactive waste

storage and handling. Additional events were

postulated for decommissioning activities

due to the difference in the types of activities

that were to be performed. The postulated

accidents described in the Defueled Safety

Analysis Report (DSAR) are generally

classified as: 1) radioactive release from a

subsystem or component, 2) fuel handling

accident, and 3) loss of spent fuel decay heat

removal capability. The postulated events

described in the Decommissioning Plan are

grouped as: 1) decontamination,

dismantlement, and materials handling

events, 2) loss of support systems (offsite

power, cooling water, and compressed air), 3)

fire and explosions, and 4) external events

(earthquake, external flooding, tornadoes,

extreme winds, volcanoes, lightning, toxic

chemical release). These types of accidents

are discussed below.

Radioactive release from a subsystem

or component involves failure of a

radioactive waste gas decay tank

(WGDT) or failure of a chemical and

volume control system holdup tank

(HUT). For a failure of a WGDT, the

radioactive contents are assumed to be

principally noble gases krypton and

xenon, the particulate daughters of some

of the krypton and xenon isotopes, and

trace quantities of halogens. For the

failure of a HUT, the assumptions were

full power operation with 1-percent

failed fuel, 40 weeks elapsed since

power operation, and 60,000 gallons of

120°F liquid released over a 2-hour

period. However, the WGDTs and HUTs

are no longer active and have been

drained. Therefore, pre-operational

testing and load handling activities

cannot increase the probability of

occurrence of a failure of a WGDT or

HUT. Since the failure of a WGDT or

HUT is no longer credible, the

consequences of failure of a WGDT or

HUT cannot significantly increase as a

result of pre-operational testing and

load handling.

The fuel handling accident involves a

stuck or dropped fuel assembly that

results in damage of the cladding of the

fuel rods in one assembly and the

release of gaseous fission products. Preoperational testing and load handling do

not involve the movement of irradiated

fuel. A dummy assembly will be used

for fit-up testing. The fuel handling

equipment will be the same as

previously analyzed with the exception

of special tools that may be used to

manipulate the dummy fuel assembly.

These special tools will be similar in

size and weight to other tools used for

underwater manipulation, and

therefore, would not present a new

hazard. In addition, the same

administrative controls and physical

limitations imposed on any fuel

handling operation will be used for preoperational testing and load handling.

Thus, there is no increase in the

probability of occurrence of a fuel

handling accident over what would be

expected for any routine fuel handling

operation. If a dummy fuel assembly

were dropped in the spent fuel pool,

then only one fuel assembly could be

damaged. Therefore, the consequences

of a dummy fuel assembly drop would

be the same as the consequences of the

analysis described in the DSAR.

Therefore, the consequences of a

dummy fuel assembly drop are not

significantly increased as a result of preoperational testing and load handling.

The loss of spent fuel decay heat

removal capability involves the loss of

forced spent fuel cooling with and

without concurrent spent fuel pool

(SFP) inventory loss. The only

requirement to assume adequate decay

heat removal capability for the spent

fuel is to maintain the water level in the

SFP so that the spent fuel assemblies

remain covered (i.e., the capability to

makeup water to the SFP must be

available when required). The potential

events that could result in a loss of

spent fuel decay heat removal capability

include external events (explosions,

toxic chemicals, fires, ship collision

with the intake structure, oil or

corrosive liquid spills in the river,

cooling tower collapse, seismic events,

severe meteorological events), and

internal events, including SFP makeup

water system malfunctions. Preoperational testing and load handling

will not require the use of explosive

materials, toxic chemicals, or flammable

materials. The probability of other

external events (e.g., cooling tower

collapse) would be unaffected by the

pre-operational testing and load

handling activities inside the fuel

building. Pre-operational testing and

load handling activities will not directly

interface with the SFP makeup water

systems, and therefore could not affect

their probability of failure. The safe load

path and handling height limitations

will ensure that a load drop does not

adversely affect the SFP or makeup

water systems. Therefore, there is no

significant increase in the probability of

a loss of spent fuel decay heat removal

capability. There are no credible adverse

consequences of the loss of spent fuel

decay heat removal as the DSAR

demonstrates that adequate time is

available to establish a source of

makeup water to the SFP such that

uncovering the fuel and an actual loss

of spent fuel cooling is not credible. The

postulated events that could affect the

SFP (liner tear/breach and heavy load

drop) do not have a significant adverse

effect. In addition, establishment of the

makeup water path and recovery of

spent fuel cooling would not be affected

because postulated off-normal events

and accidents could not affect the

capability to provide makeup water to

the SFP by various water sources.

Therefore, pre-operational testing and

load handling cannot significantly

increase the consequences of the loss of

spent fuel decay heat removal.

The events postulated in the

Decommissioning Plan are similar to the

DSAR with the exception of

decontamination, dismantlement, and

materials handling events.

Decontamination events involve gross

liquid leakage from in-situ

decontamination equipment or

accidental spraying of liquids

containing concentrated contamination.

Dismantlement events include

segmentation of components and

structures, or removal of concrete by

rock splitting, explosives, or electric

and/or pneumatic hammers.

Dismantlement events potentially result

in airborne contamination. Materials

handling events involve dropping

contaminated components, concrete

rubble, or filters or packages of

particulate materials. Pre-operational

testing and load handling activities are

material handling activities and are

therefore, within the bounds of the

existing analysis. Therefore, the

probability and consequences of

decontamination, dismantlement, and

materials handling events would not be

significantly increased.
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 79

Based on the above, the preoperational testing and load handling

activities do not involve a significant

increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes would not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. As

described in the licensee’s safety

evaluation of the proposed preoperational testing and load handling

activities, no types of off-normal events/

accidents were determined to have

radiological consequences greater than

currently evaluated in the DSAR and

Decommissioning Plan.

The postulated dummy fuel assembly

drop is considered the same type or

kind of event as the previously analyzed

fuel handling accident, mainly because

the initiator for this postulated event is

the same (i.e., a (non-specified) failure

of the fuel handling equipment or the

fuel handling bridge crane. During preoperational testing and load handling, a

dummy fuel assembly could be dropped

in the SFP or the cask loading pit. As

the cask loading pit is similar in

construction to the SFP and the cask

loading pit will be flooded with borated

water of the same concentration as the

SFP, the differences between the two

events are negligible and the two events

may be considered the same type or

kind of accident. Therefore the dummy

fuel assembly drop is not a new or

different type or kind of accident.

The postulated transfer cask drop or

mishandling event is similar to a

materials handling event. Therefore, the

consequences of a transfer cask drop or

mishandling event would not represent

a new or different type or kind of

accident.

Based on the above, the preoperational testing and load handling

activities do not create the possibility of

a new or different kind of accident.

The proposed changes do not involve

a significant reduction in the margin of

safety. The Trojan Permanently

Defueled Technical Specifications

(PDTS) contain four limiting conditions

of operation that address SFP water

level, SFP boron concentration, SFP

temperature, and SFP load restrictions.

These PDTS will remain in effect as

long as spent fuel is stored in the SFP,

which is in accordance with their

applicability statements. The preoperational testing and load handling

activities will not affect these PDTS or

their bases.

The cask loading pit (CLP) is

immediately adjacent to the SFP. The

gate between the CLP and the SFP may

be opened to allow a dummy fuel

assembly to moved from the spent fuel

storage racks in the SFP to the basket in

the CLP. Opening the gate will allow

free exchange of water between the CLP

and the SFP. The water in the CLP must

be at essentially the same level, boron

concentration, and temperature as the

SFP prior to the first opening of the gate

to ensure that the limited conditions of

operation are continuously satisfied for

the SFP. Therefore, the CLP will be

initially filled to about the same level as

the SFP with water that is about the

same boron concentration and

temperature as the SFP. With these

precautions, the limiting conditions of

operation for SFP level, boron

concentration, and temperature will be

continuously maintained and the

margin of safety will be unaffected.

Pre-operational testing and load

handling activities will involve lifting

and moving heavy loads (e.g., transfer

casks). Loads that will be carried over

fuel in the SFP racks and the heights at

which they may be carried will be

limited in accordance with LCO 3.1.4,

‘‘Spent Fuel Pool Load Restrictions,’’ in

such a way as to preclude impact

energies over 240,000 in-lbs. With this

precaution, the limiting condition of

operation pertaining to load restrictions

over the SFP will be satisfied for fuel

stored in the SFP racks and the margin

of safety will be unaffected. The safe

load path for heavy loads being lifted

and moved outside the SFP will be

located sufficiently far from the SFP as

to not have an adverse effect on the SFP

in the unlikely event of a load drop. In

addition, the mechanical stops and

electrical interlocks on the fuel building

overhead crane will provide additional

assurance that heavy loads are not

carried over the fuel in the SFP racks.

Based on the above, the preoperational testing and load handling

activities will not reduce the margin of

safety.

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: Branford Price Millar Library,

Portland State University, 934 S.W.

Harrison Street, P.O. Box 1151,

Portland, Oregon 97207

Basis for proposed no significant

hazards consideration determination:

Attorney for licensees: Leonard A.

Girard, Esq., Portland General Electric

Company, 121 S.W. Salmon Street,

Portland, Oregon 97204

NRR Project Director: Seymour H.

Weiss

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: January

2, 1997

Description of amendment request:

The proposed amendment would allow

a change to the current functional

testing frequency for Inservice

Inspection of American Society of

Mechanical Engineers Code Class 1, 2,

and 3 pumps and valves from the

current monthly to a quarterly testing

frequency.

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: Operation of Indian Point 3 in

accordance with the proposed license does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes involve no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. 10 CFR

50.55a(g) requires that safety related

components (e.g. - pumps and valves) be

tested according to the requirements of

Section XI of the American Society of

Mechanical Engineers (ASME) Boiler and

Pressure Vessel Code (Code) and applicable

addenda. The revision of functional test

frequencies for pumps and valves, which are

categorized as Code Class 1, 2, or 3, from a

monthly to a quarterly test interval is

consistent with NRC guidance provided in

NUREG-1366 and in accordance with

recommended test intervals in the ASME

Code. These changes will reduce component

degradation resulting from unnecessary tests

and provide better system availability from

not having to remove a system/component

from operability while performing a

surveillance. Such changes will not alter the

probability or consequences of any

previously analyzed accidents.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response: The proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes are procedural in

nature concerning the functional testing

frequencies of pumps and valves that have

historically shown a high percentage of

successfully meeting surveillance

requirements. The methodology of testing

these pumps and valves will remain

unchanged. The proposed changes, while

slightly increasing the possibility of an
80 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

undetected pump or valve defect, will not

create a new or unevaluated accident or

operating condition.

(3) Does the proposed license amendment

involve a significant reduction in a margin of

safety?

Response: The proposed license

amendment does not involve a significant

reduction in a margin of safety.

The proposed changes are in accordance

with recommendations provided by the NRC

regarding the improvement of Technical

Specifications. These changes will result in

the perpetuation of current safety margins

while reducing the testing burden and

decreasing equipment degradation.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa,

Acting Director

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: February

11, 1997

Description of amendment request:

The proposed change to 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,’’ would 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.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes involve: 1) no

hardware changes; 2) no significant 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. Therefore

these changes will not increase the

probability of an accident previously

evaluated.

Establishment of [Emergency Diesel

Generator] EDG fuel oil testing requirements

in TS 6.8.4.e is a change that is consistent

with changes made in the improved STS

[Standard Technical Specifications] as

contained in Specification 5.5.10 of that

document. These changes establish a new

requirement to test for particulates in the

EDG fuel oil, but establish a 92 day test

frequency (as opposed to 31 days in the

improved STS) and a 3.0 micron acceptance

criteria (as opposed to 0.8 micron in the

improved STS) for particulate testing. [Public

Service Electric and Gas Company] PSE&G

concludes that these changes are acceptable

based upon past EDG fuel oil tests for

particulates and acceptable performance of

the EDG with 5.0 micron filters. In addition,

PSE&G will utilize more objective test criteria

for water and sediment in the EDG fuel oil

than established by the ‘‘clear and bright’’

acceptance criteria contained in the

improved STS.

Since the EDG fuel oil will still: 1) meet all

of the requirements established for fuel oil

specified in the improved STS; and 2) retain

the capability to mitigate the consequences of

accidents described in the [Hope Creek

Generating Station] HC Safety Analysis

Report, the proposed changes were

determined to be justified. Based on

established fuel oil quality history, the

proposed testing methods and frequencies

will not significantly decrease confidence in

fuel oil quality and EDG operability, nor will

they have any negative effect on established

plant practices in regards to the testing of

EDG fuel oil. Therefore, these changes will

not involve a significant increase in the

consequences of an accident previously

evaluated.

The revisions proposed to the TS Bases are

being made to provide additional information

supporting the proposed EDG TS. With the

approval of the proposed TS changes, the

associated Bases changes would be editorial

in nature. Therefore, these changes will not

involve a significant increase in the

consequences of an accident previously

evaluated.

In addition, the proposed change to

[Limiting Condition for Operation] LCO

3.8.1.1, ACTION c., is considered to be

editorial in nature and will not result in 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.

The HC EDGs are designed to mitigate the

consequences of accidents by providing

electrical power to safety-related equipment.

Failure of the EDGs are not considered to

initiate any of the accidents described in the

HC Safety Analysis Report. The proposed

changes concern fuel oil system surveillances

and testing frequency. The proposed changes

will not adversely impact the operation of

any safety related component or equipment.

Since the proposed changes involve: 1) no

hardware changes; 2) no significant 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. 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.

In addition, the proposed change to LCO

3.8.1.1, ACTION c., is considered to be

editorial in nature and will not result in 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.

Establishment of EDG fuel oil testing

requirements in TS 6.8.4.e is a change that

is consistent with changes made in the

improved STS. The proposed changes

address: 1) how EDG fuel oil quality is to be

determined; 2) how frequently this

determination is to be performed; and 3) how

to control the process for determining fuel oil

acceptability and resultant EDG operability.

With the exception of particulate testing

(which is being added) all acceptance criteria

for fuel oil testing remain unchanged. Based

on historical data, EDG fuel oil quality will

not be adversely affected or impacted by the

proposed changes. Therefore, the proposed

amendment does not involve any significant

reduction in a safety margin.

The revisions proposed to the TS Bases are

being made to provide additional information

supporting the proposed EDG TS. With the

approval of the proposed TS changes, the

associated Bases changes would be editorial

in nature. Therefore, these changes will not

involve a significant reduction in a safety

margin.

In addition, the proposed change to LCO

3.8.1.1, ACTION c., is considered to be

editorial in nature and will not involve a

significant reduction in a safety margin.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket No.

50-390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: October

23, 1996, January 31, February 10 and

24 and March 11, 1997.

Description of amendment request:

The proposed amendment would revise
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 81

the Watts Bar Nuclear Plant (WBN) Unit

1 Technical Specifications to increase

the enrichment and storage capacity of

the spent fuel pool racks. The proposed

modification increases the (Watts Bar

Nuclear Plant) WBN spent fuel storage

capacity from 484 fuel assemblies to

1835 fuel assemblies. The initial

enrichment of the fuel to be stored in

the spent fuel storage racks will be

increased from 3.5 weight percent

(wt%) to 5.0 wt%. This modification

would also change the spacing of stored

fuel assembly center-to-center spacing

from a nominal 10.72 inches to 10.375

inches in 24 PaR flux trap rack modules

and 8.972 inches in ten smaller burnup

credit rack modules to be installed

peripherally along the south and west

pool walls and in a single 15 x 15

burnup credit rack to be installed in the

cask pit.

In addition to the above proposed

revisions, two limiting conditions for

operation will be added to require that

the combination of initial enrichment

and burnup of each spent fuel assembly

to be stored is in the acceptable region

and to require boron concentration of

the cask pit to be greater than or equal

to 2000 parts per million (ppm) during

fuel movement in the flooded cask pit.

As an added protection to the fuel

stored in the cask pit area, the Technical

Requirements Manual (TRM) is being

revised to require that an impact shield

be in place over the fuel when heavy

loads are moved near or across the cask

pit area.

The WBN Unit 1 Technical

Specification Bases and the TRM would

be revised to support these changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The Nuclear Regulatory Commission has

provided standards for determining whether

a significant hazards consideration exists (10

CFR 50.92(c)). A proposed amendment to an

operating license for a facility involves no

significant hazards consideration if 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.

Each standard is discussed below for the

proposed amendment.

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The following potential scenarios were

considered:

1. A spent fuel assembly drop.

2. Drop of the transfer canal gate or the

cask pit divider gate.

3. A seismic event.

4. Loss-of-cooling flow in the spent fuel

pool.

5. Installation activities.

The effect of additional spent fuel pool

storage cells fully loaded with fuel on the

first four potential accident scenarios listed

above has been considered. It was concluded

that after installation activities have been

completed, the presence of additional fuel in

the pool does not increase the probability of

occurrence of these four events. Also, based

on evaluations of bulk pool temperature, rack

seismic responses, and refueling accidents, it

is reasonable to conclude that there is no

significant increase in the consequences of

these events after installation is complete

(See Reference 1). During the installation

activities, the following considerations

support a conclusion that neither the

probability or consequences of these four

scenarios would be significantly increased.

A spent fuel assembly cannot be dropped

during installation of the 24 Programmed and

Remote System Corporation (PaR) flux trap

rack modules because this activity will take

place before the end of operating cycle one

and there will be no spent fuel in the WBN

pool to be moved or shuffled. Before

installing the ten smaller burnup credit racks

in the pool, some fuel will be moved to create

a three foot lateral free zone clearance from

stored fuel. This would involve a one-time

movement of an estimated maximum of 225

fuel assemblies, which is less that half the

fuel movements during one refueling outage.

This does not significantly increase the

probability of dropping a fuel assembly,

particularly when the many administrative

controls and physical limitations imposed on

fuel handling operations are considered. The

fuel handling system consists of equipment

and structures utilized for safely

implementing refueling operations in

accordance with requirements of General

Design Criteria 61 and 62 of 10 CFR 50,

Appendix A. The radiological dose

consequences of dropping a 5.0 wt% fuel

assembly are different from the previous

FSAR [Final Safety Analysis Report]

evaluation for the 3.5 wt% fuel assembly.

The Beta and Gamma doses decrease and the

maximum thyroid dose increase is less than

9%. Therefore, the change in calculated dose

values is insignificant and remains well

within regulatory guidelines.

It may be necessary to move the transfer

canal gate and the cask pit divider gate

between their gated and stored positions

during installation of the burnup credit

‘‘baby’’ rack modules along the south and

west walls. During rack installation, the

previously mentioned three foot lateral free

zone clearance to stored fuel would exist.

Therefore, no heavy load would be carried

directly over irradiated fuel during

installation of the racks. There are numerous

design features which comply with NUREG0612 to preclude these gates from dropping

on spent fuel. These features include design

of the lifting devices, design of the crane, and

use of written procedures. Also, the

evaluation results for a gate drop on the racks

indicates that permanent damage to a fuel

storage cell is limited to a maximum depth

of less than six inches below the top of the

rack with no effect on the subcriticality of

fuel stored in adjacent cells. Based on the

foregoing, it is reasonable to conclude that

gate handling during the installation of the

‘‘baby’’ racks would not involve a significant

increase in the probability or consequences

of an accident.

The probability of a seismic event is not

related to installation activities. The worst

consequence resulting from a seismic event

during installation activities would occur

during handling of a rack. The consequences

would be insignificant because the Auxiliary

Building crane is seismically qualified and

both handling equipment and operations

meet the criteria of NUREG-0612.

Nevertheless, if the seismic event resulted in

a rack drop, the consequences are

insignificant, i.e., localized damage to the

pool liner and a minor leak rate which would

be small in comparison to available installed

makeup capacity. The cooling and shielding

of the spent fuel would remain unaffected.

Also the racks being moved are empty during

installation and therefore, the criticality

consequences of seismic events are bounded

by evaluations for loaded racks.

Rack installation activities cannot cause an

accidental loss-of-cooling flow in the spent

fuel pool. The vital components of the spent

fuel pool cooling and cleanup system

(SFPCCS) are not located proximate to the

pool installation activities. Coolant flow may

be deliberately curtailed to facilitate

installation of the ‘‘baby’’ racks directly

beneath the discharge piping in the

southwest corner of the pool. The effects of

such an action would be readily minimized

and made inconsequential during the

detailed installation planning phase by

selecting a time when decay heat input from

stored fuel is relatively constant. Also careful

preplanning of the work would minimize

out-of-service time and provide for

intermittent coolant flow restart, if necessary,

to maintain acceptable bulk coolant

temperatures. Similarly, the effect of an

independently initiated loss-of-coolant flow

incident on reracking activities can be easily

accommodated by stopping work, as

necessary, to mitigate any adverse effects on

the installation process. The consequences of

loss-of-cooling flow in the spent fuel pool

during installation are bounded by the

analysis in Chapter 5 of the report which

includes the situation in which ‘‘baby’’ racks

and the 15 x 15 cask pit rack are installed,

and the pool is filled to capacity with spent

fuel.

With regard to the actual installation

activities, the existing WBN TRM prohibits

loads in excess of 2059 pounds from travel

over fuel assemblies in the storage pool and

requires the associated crane interlocks and

physical stops be periodically demonstrated

operable. During installation, racks and

associated handling tools will be moved over

the spent fuel pool, however there will be no

fuel in the pool when the 24 flux trap rack

modules are installed. A three foot lateral

free zone clearance from stored spent fuel
82 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

will be maintained during installation of the

ten smaller burnup credit rack modules.

Installation work in the spent fuel pit area

will be controlled and performed in strict

accordance with specific written

instructions.

NUREG-0612 states that in lieu of

providing a single failure-proof crane system,

the control-of-heavy-loads guidelines can be

satisfied by establishing that the potential for

a heavy load drop is extremely small. Storage

rack movements to be accomplished with the

WBN Auxiliary Building crane will conform

with NUREG-0612 guidelines in that the

probability of a drop of a storage rack is

extremely small. The crane has a tested

capacity of 125 tons. The maximum weight

of any existing, replacement, or new storage

rack and its associated handling tool is less

than 20 tons. Therefore, there is ample safety

factor margin for movements of the storage

racks by the Auxiliary Building crane.

Special lifting devices, which have

redundancy or a rated capacity sufficient to

maintain adequate safety factors, will also be

utilized in the movements of the storage

racks. In accordance with NUREG-0612,

Appendix B, the safety margin ensures that

the probability of a load drop is extremely

low.

Future load travel over fuel stored in a rack

specifically designed for the cask loading

area of the cask pit will be prohibited unless

an impact shield, which has been specifically

designed for this purpose, is covering the

area. Loads that are permitted when the

shield is in place must meet analytically

determined weight, travel height, and crosssectional area criteria that preclude

penetration of the shield. A Technical

Requirement (TR) has been proposed that

incorporates the previously mentioned load

criteria.

Also a rack change-out sequence is being

developed that addresses removal of the

existing racks, movement of the new racks

into the Auxiliary Building, initial staging on

the refueling floor, and final installation in

the pool. The change-out sequence objectives

include establishing lift heights, travel

distances, and number of lifts to be as low

as reasonably achievable. Accordingly, it is

concluded that the proposed installation

activities will not significantly increase the

probability of a load-handling accident. The

consequences of a load-handling accident are

unaffected by the proposed installation

activities.

The consequences of a spent fuel assembly

drop were evaluated, and it was determined

that the racks will not be distorted such that

the racks would not perform their safety

function. The criticality acceptance criterion,

Keff less than or equal to 0.95, is not violated,

and the calculated doses are well within 10

CFR Part 100 guidelines. The radiological

consequences of the fuel assembly drop

accident evaluated for WBN, have changed,

however, the changes do not involve a

significant increase in consequences and are

well within the 10 CFR 100 requirements.

A TRM change has been proposed that

would permit the transfer-canal gate and the

divider gate for the cask pit to travel over fuel

assemblies in the spent fuel pool during

movement between their gated and stored

position. Rack damage is restricted to an area

above the active fuel region, therefore,

neither criticality nor radiological concerns

exist.

The consequences of a seismic event have

been evaluated. The replacement racks are

designed and fabricated and the new racks

will be fabricated to meet the requirements

of applicable portions of the NRC regulatory

guides and published standards. Design

margins have been provided for rack tilting,

deflection, and movement such that the racks

do not impact each other or the spent fuel

pool walls in the active fuel region during the

postulated seismic events. The free-standing

racks will maintain their integrity during and

after a seismic event. The fuel assemblies

also remain intact and therefore no criticality

concerns exist.

The spent fuel pool system is a passive

system with the exception of the fuel pool

cooling train and heating, ventilating, and

air-conditioning (HVAC) equipment.

Redundancies in the cooling train and HVAC

hardware are not reduced by the planned fuel

storage modification. The potential increased

heat load resulting from any additional

storage of spent fuel is well within the

existing system cooling capacity. Therefore,

the probability of occurrence or malfunction

of safety equipment leading to the loss-ofcooling flow in the spent fuel pool is not

significantly affected. Furthermore, the

consequences of this type incident are not

significantly increased from previously

evaluated cooling system loss of flow

malfunctions. Thermal-hydraulic scenarios

assume the reracked pool is approximately

90% full with spent fuel assemblies. From

this starting point, the remaining storage

capacity is utilized by analyzing both normal

and unplanned full core off loads using

conservative assumptions and previously

established methods. Calculated values

include maximum pool water bulk

temperature, coincident maximum pool

water local temperature, the maximum fuel

cladding temperature, time-to-boil after lossof-cooling paths, and the effect of flow

blockage in a storage cell.

Although the proposed modification

increases the pool heat load, results from the

above analyses yield a maximum bulk

temperature less than 160 degrees Fahrenheit

which is below the bulk boiling temperature.

Also the maximum local water temperature

is below nucleate boiling condition values.

Associated results from corresponding lossof-cooling evaluations give minimums of 5.3

hours before boiling begins and 45 hours

before the pool water level drops to the

minimum required for shielding spent fuel.

This is sufficient time to begin utilization of

available alternate sources of makeup cooling

water. Also, the effect of the increased

thermal loading on the pool structure,

associated cooling system, and components

was evaluated and determined to establish an

acceptable design basis with the new storage

configuration. No modifications were

necessary because of the increased

temperature.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed modification has been

evaluated in accordance with the guidance of

the NRC position paper entitled, ‘‘OT

Position for Review and Acceptance of

Spent-Fuel Storage and Handling

Applications’’, appropriate NRC regulatory

guidelines; appropriate NRC standard review

plans; and appropriate industry codes and

standards. Proven analytical technology was

used in designing the planned fuel storage

expansion and will be utilized in the

installation process. Basic reracking

technology has been developed and

demonstrated in applications for fuel pool

capacity increases that have already received

NRC staff approval.

Proposed TSs for the spent fuel storage

racks use burnup credit and fuel assembly

administrative placement restrictions for

criticality control. These restrictions are

described in the proposed change to the

design features section of the TSs by

reference to the Spent Fuel Pool

Modifications report. Additional evaluations

were required to ensure that the criticality

criterion, keff less than 0.95, is maintained.

These include evaluation for the abnormal

placement of unirradiated (fresh) fuel

assemblies of 5.0 wt% enrichment into a

storage cell location designed for lower

enrichment or irradiated fuel. Soluble boron,

for which credit is permitted under these

abnormal conditions, ensures that reactivity

is maintained substantially less than the

design requirement. For example, if the PaR

flux trap racks are inadvertently all loaded

with fresh assemblies of the maximum 5.0

wt% fuel instead of observing the 3.8 wt%

and 6.75 MWD/KgU controls, the worth of

the 2000 ppm borated water is sufficient to

lower the keff of the storage racks to 0.83. The

existing and proposed TSs require boron

concentration in the pool and cask pit to be

more than or equal to 2000 ppm during fuel

movement. An analytical determination of

the reactivity worth of 2000 ppm borated

water in the spent fuel storage pool predicted

the change in keff to be approximately 17

percent keff. Although no credit for soluble

boron was proposed in the TSs, it was also

determined by an independent calculation

that a minimum concentration of 520 ppm

soluble boron allows the unrestricted storage

of 5.0 wt% enriched fuel in the PaR flux trap

racks.

The Holtec-designed peripheral ‘‘baby’’

racks and the 15 x 15 racks in the cask

loading area can safely and conservatively

store fuel of 5 wt% initial enrichment burned

to 41 MWD/kgU or lower enriched fuel with

lower burnup, i.e., fuel of equivalent

reactivity. Evaluations have confirmed that,

for the abnormal placement of a fresh fuel

assembly of 5.0 wt% in these racks, the

criticality criterion is maintained with the

existing and proposed TS requirements of

2000 ppm soluble boron.

Although these changes required

addressing additional aspects of a previously

analyzed accident, the possibility of a

previously unanalyzed accident is not

created.

The impact shield design together with its

attendant administrative controls and

NUREG-0612 heavy load lift compliance,

renders the possibility of a heavy load drop
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 83

on fuel as not credible in accordance with the

NUREG-0612 single-failure-proof criteria.

Accordingly, since this particular part of the

proposed reracking modification is not a

change that could malfunction by a new

single failure, the movement of heavy loads

over the cask pit does not create the

possibility of a new or different kind of

accident.

It is therefore concluded that the proposed

reracking does not create the possibility of a

new or different kind of accident from any

previously analyzed.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The design and technical review process

applied to the reracking modification

included addressing the following areas:

1.

Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material, and structural

considerations.

The established acceptance criterion for

criticality is that the neutron multiplication

factor shall be less than or equal to 0.95,

including all uncertainties. The results of the

criticality analyses for the rack designs

demonstrate that this criterion is satisfied.

The methods used in the criticality analysis

conform to the applicable portions of NRC

guidance and industry codes, standards, and

specifications. In meeting the acceptance

criteria for criticality in the spent fuel pool

and the cask loading area, such that keff is

always less than 0.95 at a 95/95 percent

probability tolerance level, the proposed

amendment does not involve a significant

reduction in the margin of safety for nuclear

criticality.

Conservative methods and assumptions

were used to calculate the maximum fuel

temperature and the increase in temperature

of the water in the spent fuel pit area. The

thermal-hydraulic evaluation used methods

previously employed. The proposed storage

modification will increase the heat load in

the spent fuel pool, but the evaluation shows

that the existing spent fuel cooling system

will maintain the bulk pool water

temperature at or below 160 degrees

Fahrenheit. Thus it is demonstrated that the

worst-case peak value of the pool bulk

temperature is considerably lower than the

bulk boiling temperature. Evaluation also

shows that maximum local water

temperatures along the hottest fuel assembly

are below the nucleate boiling condition

value. Thus, there is no significant reduction

in the margin of safety for thermal hydraulic

or spent fuel cooling considerations.

The mechanical, material, and structural

design of the spent fuel racks is in

accordance with applicable portions of

NRC—s position in ‘‘OT Position for Review

and Acceptance of Spent-Fuel Storage and

Handling applications,’’ dated April 14, 1978

(as modified January 18, 1979), as well as

other applicable NRC guidance and industry

codes. The primary safety function of the

spent fuel racks is to maintain the fuel

assemblies in a safe configuration through

normal and abnormal loading conditions.

Abnormal loadings that have been evaluated

with acceptable results and discussed

previously include the effect of an

earthquake and the impact because of the

drop of a fuel assembly. The rack materials

used are compatible with the fuel assemblies

and the environment in the spent fuel pool.

The structural design for the new racks

provides tilting, deflection, and movement

margins such that the racks do not impact

each other or the spent fuel pit walls in the

active fuel region during the postulated

seismic events. Also the spent fuel

assemblies themselves remain intact and no

criticality concerns exist. In addition, finite

element analysis methods were used to

evaluate the continued structural

acceptability of the spent fuel pit. The

analysis was performed in accordance with

‘‘Building Code Requirements for Reinforced

Concrete,’’ (ACI 318-63,77). Therefore, with

respect to mechanical, material, and

structural considerations, there is no

significant reduction in a margin of safety.

Summary

Based on the above analysis, TVA has

determined that operation of WBN, in

accordance with the proposed amendment,

would not: (1) involve a significant increase

in the probability of consequences of an

accident previously evaluated, (2) create the

possibility of a new or different kind of

accident from any accident previously

evaluated, or (3) involve a significant

reduction in a margin of safety. Therefore,

operations of WBN in accordance with the

proposed amendments as described do not

involve significant hazard considerations as

defined in 10 CFR 50.92 and that the criteria

of 10 CFR 50.91 have accordingly been met.

TVA has also reviewed the NRC examples

of licensing amendments considered not

likely to involve significant hazards

considerations as provided in the final

adoption of 10 CFR 50.92 published on page

7751 of the Federal Register, Volume 51, No.

44, March 6, 1986. Example (X) provides four

criteria that, if satisfied by a reracking

request, indicate that it is likely no

significant hazards considerations are

involved. The criteria and how TVA—s

amendment request for WBN complies are

indicated below.

Criterion (1):

The storage expansion method consists of

either replacing existing racks with a design

that allows closer spacing between stored

spent fuel assemblies or replacing additional

racks of the original design on the pool floor

if space permits.

Proposed Amendment:

The WBN reracking involves replacing the

existing racks with a design that allows

slightly closer spacing between stored fuel

assemblies and also provides additional rack

storage on the pool floor where space

permits.

Criterion (2):

The storage expansion method does not

involve rod consolidation or double tiering.

Proposed Amendment:

The WBN racks are not double tiered, and

the racks will sit on the floor of the spent fuel

pool. Additionally, the amendment

application does not involve consolidation of

spent fuel.

Criterion (3):

The keff of the pool is maintained less than

or equal to 0.95.

Proposed Amendment

The design of the spent fuel racks contains

a neutron absorber, Boral, to allow close

storage of spent fuel assemblies while

ensuring that the keff remains less than 0.95

under normal operating conditions with

unborated water in the pool and less than

0.95 under abnormal conditions with soluble

boron in the pool.

Criterion (4):

No new technology or unproven

technology is utilized in either the

construction process or the analytical

techniques necessary to justify the

expansion.

Proposed Amendment:

The construction processes and analytical

techniques used in the fabrication and design

are substantially the same as those of

numerous other rack installations, Thus, no

new or unproven technology is utilized in

the construction or analysis of the high

density, spent fuel racks at WBN. TVA’s

contractor, Holtec International, has

previously supplied licensable racks of

several similar design for about 10 other

reracking projects

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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Virginia Electric and Power Company,

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

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request:

November 26, 1996

Description of amendment request:

The proposed changes would eliminate

the records retention requirements from

the administrative section of the

Technical Specifications (TS) in

accordance with NRC Administrative

Letter 95-06, ‘‘Relocation of Technical

Specifications Administrative Controls

Related to Quality Assurance.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Specifically, operation of the Surry...

Power [Station] in accordance with the
84 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

proposed Technical Specifications changes

will not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed

administrative changes do not affect

equipment or its operation. Therefore, the

likelihood that an accident will occur is

neither increase nor decreased by relocating

record retention requirements from the

Technical Specifications to the Operational

Quality Assurance Program. This TS change

will not impact the function or method of

operation of plant equipment. Thus, a

significant increase in the probability of a

previously analyzed accident does not result

due to this change. No systems, equipment,

or components are affected by the proposed

changes. Thus, the consequences of any

accident previously evaluated in the UFSAR

[Updated Final Safety Analysis Report] are

not increased by this change.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated. The proposed change

does not alter the design or operations of the

physical plant. Since record retention

requirements are administrative in nature, a

change to these requirements does not

contribute to accident initiation, an

administrative change related to this activity

does not produce a new accident scenario or

produce a new type of equipment

malfunction. [These] changes do not alter any

existing accident scenarios. The proposed

administrative change does not affect

equipment or its operation, and, thus, does

not create the possibility of a new or different

kind of accident. Therefore, the proposed

change does not create the possibility of a

new or different kind of accident.

(3) Involve a significant reduction in a

margin of safety. Section 6.0 of the...Surry

Technical Specifications does not have a

basis description. The proposed

administrative change does not affect

equipment or its operation, and, thus, does

not involve any 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: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: Mark Reinhart,

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.

Baltimore Gas and Electric Company,

Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2,

Calvert County, Maryland

Date of application for amendment:

July 31, 1997, as supplemented

February 13, 1997.

Brief description of amendment: The

proposed amendment would revise the

Technical Specifications to reduce the

minimum Reactor Coolant System total

flow rate from 370,000 gpm to 340,000

gpm. The proposed changes are

necessary to support a larger number of

plugged steam generator tubes for future

operating cycles.

Date of publication of individual

notice in Federal Register: February 26,

1997 (62 FR 8780)

Expiration date of individual notice:

March 28, 1997

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Consolidated Edison Company of New

York, Docket No. 50-247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

February 14, 1997

Brief description of amendment: The

proposed amendment would revise the

Technical Specifications to permit a

one-time extension of the current steam

generator tube inservice inspection

cycle. Date of publication of individual

notice in Federal Register: March 4,

1997 (62 FR 9816)

Expiration date of individual notice:

March 28, 1997

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

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

Brief description of amendment:

Changes to Technical Specification to

implement 10 CFR 50, Appendix J

Option B relating to containment

leakage tests.

Date of publication of individual

notice in the Federal Register: February

28, 1997 (62 FR 9214).

Expiration date of individual notice:

March 31, 1997

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: February

14, 1997

Brief description of amendment

request: The proposed amendment

would revise Technical Specification

(TS) Section 3/4.5.2, ‘‘Emergency Core

Cooling Systems, ECCS Subsystems -

Tavg more than or equal to 280°F.’’

Surveillance requirement 4.5.2.f would

be modified to state that opening and

closing of the inspection port on the

watertight enclosure for the decay heat

valve pit would not require this

surveillance procedure to be performed.

The applicable TS bases would also be

changed. Date of publication of

individual notice in Federal Register:

February 26, 1997 (62 FR 8783)

Expiration date of individual notice:

March 28, 1997

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

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
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 85

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

August 1, 1996

Brief description of amendments: The

amendments modify the Technical

Specifications requirements to allow use

of blind flanges during Modes 1-4 in the

Calvert Cliffs 1 and 2 Containment

Purge system instead of the two

outboard 48-inch isolation valves. Date

of issuance: March 7, 1997

Effective date: As of the date of

issuance to be implemented by the end

of the 1998 refueling outage for Unit 1;

by the end of the 1997 refueling outage

for Unit 2.

Amendment Nos.: 221 and 197

Facility Operating License Nos. DPR53 and DPR-69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47975) The Commission’s related

evaluation of these amendments is

contained in a Safety Evaluation dated

March 7, 1997 No significant hazards

consideration comments received: No

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

December 30, 1996

Brief description of amendment: The

amendment revises chemistry data for

TS Figures 3.4-2 and 3.4-3 and the

associated Bases.

Date of issuance: March 7, 1997

Effective date: March 7, 1997

Amendment No.: 68

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4342)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 7, 1997. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

September 20, 1996, as supplemented

January 21, 1997.

Brief description of amendments: The

amendments would update the

pressure- temperature cures contained

in the Dresden and Quad Cities

Technical Specifications to 22 Effective

Full Power Years. Date of issuance:

February 28, 1997 Effective date:

Immediately, to be implemented within

30 days.

Amendment Nos.: 153, 148, 172 and

168

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66703). The January 21, 1997, submittal

provided additional clarifying

information that did not change the

original proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated February 28, 1997 No

significant hazards consideration

comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

December 6, 1996

Brief description of amendments: The

amendments would change the

Technical Specification (TS) by

allowing a single control rod to be

moved when the plant is in the Hot

Shutdown or Cold Shutdown condition

provided that the one-rod-out interlock

is Operable and the reactor mode switch

is in the refuel position.

Date of issuance: March 4, 1997

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 154, 149, 173, 169

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2187).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 4, 1997. No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

January 6, 1997

Brief description of amendments: The

amendments would change the

technical specifications to clarify and

maintain consistency between the

operability requirements for protective

instrumentation and associated

automatic bypass features.

Date of issuance: March 14, 1997

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 155, 150, 174, 170
86 Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6573). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 14, 1997. No significant hazards

consideration comments received: No

Local Public Document Room

location: For Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request:

November 26, 1996 as supplemented by

letters dated December 17, 1996, March

4, 1997, and March 10, 1997

Brief description of amendment: The

amendment changes reactor coolant

systems pressure/temperature limits to

incorporate updated parameters and

requirements.

Date of issuance: March 14, 1997

Effective date: March 14, 1997

Amendment No.: 188

Facility Operating License No. DPR51. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4346)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 14, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

April 11, 1996 as supplemented by

letters dated June 18, and September 5,

1996.

Brief description of amendment: The

amendment adds low-temperature

overpressure protection requirements to

the Technical Specifications as

proposed by Generic Letter 90-06.

Date of issuance: March 7, 1997

Effective date: March 7, 1997, to be

implemented within 30 days.

Amendment No.: 180

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20846) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated March 7, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

August 23, 1996, as supplemented

January 8, 1997 (TSCR 245)

Brief description of amendment: The

amendment updates the pressuretemperature limits up to 22, 27, and 32

effective full power years.

Date of Issuance: March 6, 1997

Effective date: March 6, 1997, to be

implemented within 30 days of issuance

Amendment No.: 188

Facility Operating License No. DPR16. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47977). The January 8, 1997, letter

provided clarifying information within

the scope of the original application and

did not change the staff’s initial

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of this

amendment is contained in a Safety

Evaluation dated March 6, 1997 No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

October 17, 1996, as supplemented and

modified on December 13, 1996

Brief description of amendment: The

amendment revises the Operating

License to reflect the transfer of Soyland

Power Cooperative’s 13.21-percent

minority ownership of Clinton Power

Station to Illinois Power Company. The

Operating License has been revised to

delete Soyland Power Cooperative as an

owner.

Date of issuance: March 13, 1997

Effective date: March 13, 1997

Amendment No.: 114

Facility Operating License No. NPF62: The amendment revised the

Operating License.

Date of initial notice in Federal

Register: November 19, 1996 (61 FR

58897) and January 29, 1997 (62 FR

4337) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 13, 1997. No significant hazards

consideration comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Indiana Michigan Power Company,

Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien

County, Michigan

Date of application for amendment:

June 19, 1996, and supplemented

September 19, 1996, and December 20,

1996.

Brief description of amendment: The

amendment revises the TS to allow a

permanent extension of the interim

steam generator tube voltage-based

repair criteria for steam generator tubes

used in Cycles 13, 14 and 15 at the

Donald C. Cook Nuclear Power Plant,

Unit 1.

Date of issuance: March 13, 1997

Effective date: March 13, 1997, with

full implementation within 45 days

Amendment No.: 215

Facility Operating License No. DPR58. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 31, 1996 (61 FR 40022)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 13, 1997. No

significant hazards consideration

comments received: No. The September

19, 1996, and December 20, 1996, letters

provided additional information within

the scope of the original application and

did not change the initial proposed no

significant hazards consideration

determination.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Indiana Michigan Power Company,

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

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

May 26, 1995, and supplemented

September 26, 1995, August 2, 1996 and

February 6, 1997

Brief description of amendments: The

amendments revise the TS to allow

operation of Cook Unit 1 at steam

generator tube plugging levels up to

30%. Additional changes to increase

operating margins for both Unit 1 and

Unit 2 are also included.

Date of issuance: March 13, 1997

Effective date: March 13, 1997, with

full implementation within 45 days
Federal Register / Vol. 62, No. 58 / Wednesday, March 26, 1997 / Notices 87

Amendment Nos.: 214 and 199

Facility Operating License Nos. DPR58 and DPR-74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37095)

The September 26, 1995, August 2,

1996, and February 6, 1997,

supplements provided clarifying

information that did not expand the

scope of the initial application or

change the staff’s proposed no

significant hazards determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated March 13, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: October

17, 1996

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications relating to incore detector system, seismic

instrumentation, meteorological

instrumentation, and turbine overspeed

protection. The amendment deletes

Limiting Conditions for Operation and

Surveillance Requirements related to

these instruments. The deleted

requirements are to be incorporated into

the Seabrook Station Technical

Requirements Manual (SSTR). The

associated Bases Sections are also

deleted. In addition, Technical

Specification 5.5 is deleted but will not

be relocated to the SSTR. The

amendment also redesignates Paragraph

2.J of the Seabrook Operating License as

Paragraph 3, and has added new

Paragraph 2.J to document the North

Atlantic commitment to relocate the

above mentioned Technical

Specification requirements to the SSTR.

Date of issuance: March 12, 1997

Effective date: March 12, 1997

Amendment No.: 50

Facility Operating License No. NPF86. Amendment revised the Technical

Specifications and Operating License.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66713). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 12, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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:

May 23, 1996, as supplemented July 17

and December 4, 1996

Brief description of amendment: The

amendment modifies the description of

the time constants associated with the

Overtemperature Delta-T and

Overpower Delta-T calculations used to

establish the trip setpoints and the time

constant used in the rate-lag controller

for Steam Line Isolation, Steam Line

Pressure Negative Rate-High.

Date of issuance: March 11, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 134

Facility Operating License No. NPF49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 17, 1996 (61 FR 30639)

The July 17 and December 4, 1996,

letters provided additional, clarifying

information that did not change the

scope of the May 23, 1996, application

and the initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 11, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385

Southern California Edison Company,

et al., Docket No. 50-362, San Onofre

Nuclear Generating Station, Unit No. 3,

San Diego County, California

Date of application for amendment:

February 7, 1997

Brief description of amendment: The

amendment defers implementation of

Surveillance Requirement 3.1.5.4 of

Technical Specification 3.1.5, ‘‘Control

Element Assembly (CEA) Alignment,’’

until the next SONGS Unit 3 shutdown,

which will be no later than the

upcoming Cycle 9 refueling outage

(currently scheduled for April 12, 1997).

Date of issuance: March 5, 1997

Effective date: March 5, 1997

Amendment No.: 126

Facility Operating License No. NPF15: The amendments revised the

Technical Specifications. Public

comments requested as to proposed no

significant hazards consideration: Yes

(62 FR 7477 dated February 19, 1997).

The notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by March 21, 1997,

but indicated that if the Commission

makes a final no significant hazards

consideration determination any such

hearing would take place after issuance

of the amendment. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

March 5, 1997.

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713

Dated at Rockville, Maryland, this 19th day

of March 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects

III/IV, Office of Nuclear Reactor Regulation

[Doc. 97–7508 Filed 3–25–97; 8:45 am]

BILLING CODE 7590–01–F

[Docket No. 50–409]

Lacrosse Boiling Water Reactor; Intent

To Relocate Local Public Document

Room

Notice is hereby by given that the

Nuclear Regulatory Commission (NRC)

will be relocating the local public

document room (LPDR) for records

pertaining to Dairyland Power

Cooperative’s LaCrosse Boiling Water

Reactor located in Genoa, Wisconsin.

The LaCrosse LPDR is currently located

at the LaCrosse Public Library, 800 Main

Street, LaCrosse, Wisconsin. The

document collection is available in

microfiche form, with paper copy

indices. Library staff informed the NRC

that they are no longer able to maintain

the document collection and request

that it be moved. This notice invites

public comment on possible LPDR

locations in the Genoa, Wisconsin, area.

Among the factors the NRC will

consider in selecting a new location for

the LPDR are the following:

(1) Whether the institution is an

established document repository located

near the nuclear facility with a history

of impartially serving the public;

(2) The physical facilities available,

including shelf space, storage space,

patron workspace, copying equipment

and computer access;