Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17223

Broad Street, Walhalla, South Carolina

for the Oconee Nuclear Station; the York

County Library, 138 East Black Street,

Rock Hill, South Carolina 29730 for the

Catawba Nuclear Station; and the J.

Murrey Atkins Library, University of

North Carolina at Charlotte, 9201

University City Boulevard, North

Carolina 28223 for the McGuire Nuclear

Station.

Dated at Rockville, Maryland, this 3rd day

of April 1997.

For The Nuclear Regulatory Commission.

Herbert N. Berkow,

Director, Project Directorate II–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 97–9059 Filed 4–8–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving no Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

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 17,

1997 through March 28, 1997. The last

biweekly notice was published on

March 26, 1997 (62 FR 14457).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed no Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

Involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By May 9, 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
17224 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

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: March

5, 1997.

Description of amendments request:

The proposed amendments would

incorporate a new Technical

Specification (TS) for instrumentation

associated with automatic isolation of a

pathway for release of non-condensible

gases from the main condenser. At

power levels of 5 percent or less,

mechanical vacuum pumps are used to

remove non-condensible gases from the

condenser using a pathway to the

release stack that bypasses the normal

holdup and filter train. The proposed

TS will require that four channels of the

main steam line radiation—high

isolation function be capable of tripping

the mechanical vacuum pumps and

closing an isolation valve in the release

pathway. Surveillance requirements are

included in the TS to ensure the

isolation instrumentation will perform

its intended function.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendments do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change incorporates a new

Technical Specification 3/4.3.8, ‘‘Condenser

Vacuum Pump Isolation Instrumentation.’’

This specification will require that the main

steam line radiation—high isolation function

be capable of tripping the condenser vacuum

pump(s) and isolate the associated common

isolation valve. Four instrumentation

channels of this function are required to be

operable when the unit is in OPERATIONAL

CONDITION 1 or 2 with a condenser vacuum

pump in operation. Adding the requirement

to trip the condenser vacuum pumps does

not affect the probability of an accident

previously evaluated. The probability of

component failure of the proposed design for

condenser vacuum pump isolation devices is

the same as that of the original licensing

basis. As a result, the capability to isolate the

condenser vacuum pump will not be

significantly impacted.

CP&L contracted Scientech-NUS to

recalculate the main control room doses

resulting from a control rod drop accident

assuming main steam line radiation monitors

isolate the condenser vacuum pump(s) and

determined the dose to be 23.2 rem thyroid

and 0.05 rem whole body, which is less than

the General Design Criterion (GDC) 19/

Standard Review Plan (SRP) Section 6.4

limits of 30 rem thyroid and 5 rem whole

body. The offsite doses at the exclusion area

boundary after 2 hours are 0.16 rem thyroid

and 0.015 rem whole body, which is less

than the SRP Section 15.4.9 limits. The low

population zone (LPZ) dose is estimated to be

about 1 rem thyroid, which is also well

below regulatory limits. Therefore, the

proposed [amendments do] not increase 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 [amendments add] new

requirements to ensure the capability to trip

the condenser vacuum pump(s). The

proposed [changes do] not affect the

operability of equipment designed to mitigate

the consequences of an accident nor [do they]

create a potential to initiate a new type of

accident. 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 license [amendments do]

not involve a significant reduction in a

margin of safety.

The safety-related main steam line

radiation monitors provide a highly reliable

means to detect radioactivity resulting from

a control rod drop accident and will provide

automatic trip of the condenser vacuum

pumps and isolation of the associated

isolation valve. Use of the main steam line

radiation monitors for this application is

consistent with the original Brunswick Steam

Electric Plant design for condenser pump and

associated valve isolation. CP&L contracted

Scientech-NUS to recalculate the main

control room doses resulting from a control

rod drop accident assuming main steam line

radiation monitors isolate the condenser
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17225

vacuum pump(s) and determined it to be 23.2

rem thyroid and 0.05 rem whole body, which

is less than the GDC 19/SRP Section 6.4

limits of 30 rem thyroid and 5 rem whole

body. The offsite doses at the exclusion area

boundary after 2 hours are 0.16 rem thyroid

and 0.015 rem whole body, which is less

than the SRP Section 15.4.9 limits. LPZ dose

is estimated to be about 1 rem thyroid, which

is also well below regulatory limits.

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 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: February

18, 1997.

Description of amendment request:

The proposed change revises the Plant

System Turbine Cycle Technical

Specification (TS) 3/4.7.1 by revising

the power range high neutron flux

setpoint values in TS Table 3.7–1.

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 proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The high flux setpoints are being revised

to provide additional margin against

secondary side overpressurization for LOL/

TT [loss-of-load/turbine trip] events. The

proposed revision will not create any loss or

reduction in redundancy or diversity in the

reactor protection systems that would

increase the probability of a previously

evaluated accident. The high flux setpoints

are being revised to ensure that the

consequences of a previously evaluated

accident do not increase.

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.

No new or previously unanticipated failure

mechanisms are introduced by the proposed

change. No new failure modes have been

created by the proposed change. No new

credible event or initiating factor is

introduced. Reactor power is limited to

ensure that the secondary system is not

overpressurized.

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.

The margin of safety as defined in the basis

of the Technical Specification does not

decrease. This change is proposed to ensure

that the secondary system pressure will be

limited to within 110% of its design pressure

during the most severe anticipated

operational transient. The revised high flux

setpoints are intended to bound the

allowable operating configurations of TS

Table 3.7–1.

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.

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: February

21, 1997.

Description of amendment request:

The proposed change adds a definitive

time limit to Technical Specification

3.3.2 in Action 16 of Table 3.3–3 to

place an inoperable channel into

bypass.

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 change does not affect the

operation or design of the plant in any way.

The requirement to place the channel into

bypass already exists and this change simply

provides a specific time limit. This logic

circuit is not an initiator of any event and

with no change in logic or operation there is

no change in consequences.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed specific time limit does not

involve any physical alterations or additions

to plant equipment or alter the manner in

which any safety-related system performs its

function. 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.

The proposed change replaces an

indeterminate time period with a specific

limit of six hours. Six hours is a reasonable

period in which to complete this requirement

and is identical to the time allowed for these

functions in NUREG–1431 [Standard

Specifications Westinghouse Plants].

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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.

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: February

27, 1997.

Description of amendment request:

The proposed change adds sleeve

installation as an alternative to tube

plugging for repairing degraded steam

generator tubes to Technical

Specification 3/4.4.5, Steam Generators.
17226 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

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 only equipment affected by sleeving is

the steam generator tubes. The most severe

malfunction of a steam generator tube is a

tube rupture. The consequences of a ruptured

sleeve are no greater than the consequences

of a ruptured tube. Sleeving does not increase

the probability of a steam generator tube

failure because the sleeved tube has been

shown to have a significant safety factor for

burst and collapse pressures as well as

demonstrated acceptable resistance to

corrosion and fatigue loading. Thus, a steam

generator with sleeved tubes would perform

in the same manner as one without sleeved

tubes.

A sleeved tube is functionally equivalent to

an unsleeved tube except for less effective

heat transfer due to the air gap and a slightly

higher pressure drop due to the primary flow

restriction. These differences are bounded by

the current tube plugging limits.

Analysis and testing have demonstrated

that the sleeves are structurally adequate to

withstand the load existing within the steam

generator tubes whether the original tube is

still intact or is breeched.

There is no increase in the possibility for

increased fatigue loadings. There is no

possibility for the sleeve to become dislodged

from its plugging location and enter the RCS

[Reactor Coolant System] flow path.

The plant safety analysis for tube plugging

bounds tube sleeving.

The proposed change has no significant

effect on the configuration of the plant. The

proposed change does not affect the way in

which the plant is operated. 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.

A sleeved tube is functionally equivalent to

an unsleeved tube except for less effective

heat transfer due to the air gap and a slightly

higher pressure drop due to the primary flow

restriction. These differences are bounded by

the current tube plugging limits.

The sleeved tube has been shown to have

a significant safety factor for burst and

collapse pressures as well as demonstrated

acceptable resistance to corrosion and fatigue

loading. Thus, a steam generator with sleeved

tubes would perform in the same manner as

one without sleeved tubes.

The proposed change has no significant

effect on the configuration of the plant. The

proposed change does not affect the way in

which the plant is operated. 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.

The proposed revision to permit the

installation of tube sleeves does not reduce

the margin of safety as presently defined in

Technical Specification BASES section 3/

4.4.5. This margin of safety includes primary

to secondary leakage limits and tube plugging

limits which are not changed by the

proposed amendment. The analyses and

testing of the proposed sleeve design

demonstrates that the structural integrity of

the RCS is maintained. Design of the tube

sleeve considers mechanical/structural

aspects, water chemistry and metallurgical

aspects as well as thermal/hydraulic

considerations.

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.

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

10, 1997.

Description of amendment request:

The proposed changes to Technical

Specification 3.5.1 provide an optional

method of meeting surveillance

requirements by allowing the use of

instrument readings to meet

surveillance 4.5.1.1.a.1, and adds a new

Action c to cover a condition in which

one accumulator has a boron

concentration not within limits.

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 accumulators are not initiators of any

event and so the probability of occurrence of

an event is unaffected by either of the

proposed changes. The use of actual

instrumentation readings to comply with the

surveillance does not change the function or

performance of the accumulators and thus

does not affect any accident consequences.

The increase in the allowed time to restore

the boron concentration to within limits is

consistent with allowed out of service times

for other Emergency Safeguards equipment.

It will not have a significant impact on

subcriticality during reflood. Therefore, there

will be no increase in the consequences of an

accident.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes to the accumulator

specification do not involve any physical

alterations or additions to plant equipment or

alter the manner in which any safety-related

system performs its function. 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.

The proposed change to the surveillance

requirement provides an equivalent means of

meeting the requirement. Since there is no

change in either the accumulator limits or the

surveillance frequency, there is no reduction

in safety margin. The new Action c to

address returning the boron concentration of

a single accumulator to within limits allows

an out of service time commensurate with the

times allowed for other Engineered

Safeguards Features. The boron

concentration of one accumulator does not

have a significant impact on subcriticality

during reflood and thus does not involve a

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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.

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 amendment will revise the Final

Safety Analysis Report to include the
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17227

evaluation of a spent fuel cask drop

analysis.

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 changes described do not impact the

probability of occurrence of accidents

previously analyzed. Removal of the valve

box covers and all but four of the cask

closure head sleeve nuts has no impact on

accident initiators. Dose assessments using

maximum potential releases assuming failure

of the spent fuel and radionuclide release

through the gap between the cask closure

head and the cask or damage to the valves

show that no significant increase in

consequences of an accident previously

evaluated would occur. [Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.]

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Compromising the integrity of the cask by

removing the valve box covers and closure

head sleeve nuts in preparation for unloading

the spent fuel from the cask does not create

the possibility of a new type of accident or

equipment malfunction. No safety-related

equipment, safety function, or operations of

plant equipment will be altered as a result of

this change. 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 the margin of

safety.

The NRC basis for acceptance of a spent

fuel cask drop is documented in Section

15.7.5 of the Safety Evaluation Report,

NUREG–1038, dated November 1983. It

states, ‘‘* * * no loss of cask integrity is

postulated to occur in the event of a drop,

and the staff concludes there will be no

significant radiation released to the

environment. The radiological consequences

will be less than a small fraction of the 10

CFR 100 exposure guideline values.’’

As described in the proposed change, even

though complete cask integrity may not be

preserved in the event of a loaded cask drop

with the valve box covers removed or with

only four, rather than 32, closure head sleeve

nuts installed, the radiological consequences

calculated using conservative assumptions

were determined to be a small fraction of the

10 CFR 100 values. 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. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of amendment request: June 20,

1996, as supplemented by letters dated

December 30, 1996, and March 5, 1997.

Description of amendment request:

The proposed amendments would

change the Technical Specifications

(TS) by incorporating NRC approved

thermal limit licensing methodology in

the list of approved methodologies used

in establishing the fuel cycle specific

thermal limits. In addition, the

proposed amendment would correct

mirror editorial items in the TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The probability of an evaluated accident is

derived from the probabilities of the

individual precursors to that accident. The

consequences of an evaluated accident are

determined by the operability of plant

systems designed to mitigate those

consequences. Limits will be established

consistent with NRC approved methods to

ensure that fuel performance during normal,

transient and accident conditions is

acceptable. The proposed Technical

Specifications amendment reflects NRC

approved SPC methodology used to analyze

normal operations, including anticipated

operational occurrences (AOOs), and to

determine the potential consequences of

accidents.

Licensing Methods and Models

The proposed amendment is to support

operation with NRC approved fuel and

licensing methods supplied from Siemens

Power Corporation [SPC]. In accordance with

[Updated Final Safety Analysis Report]

UFSAR Chapter 15, the same accidents and

transients will be analyzed with the new fuel

and methods. The latest NRC approved

revision to the Siemens [loss-of-coolant

accident] LOCA analysis methodology

(Reference: ANF–91–048(P)(A), Advanced

Nuclear Fuels Corporation Methodology for

Boiling Water Reactors EXEM BWR

Evaluation Model) will be used to evaluate

the ATRIUM–9B and other co-resident fuel

types. The other licensing analysis methods

and models are also NRC approved. The

approved methods and models are used to

determine the fuel thermal limits (e.g.,

average planar linear heat generation rate,

transient linear heat generation rate,

minimum critical power ratio and linear heat

generation rate). The SPC core monitoring

code enables the site to monitor keff as well

as control rod density to perform the

reactivity anomaly surveillance. Therefore,

the change in licensing analysis methods and

models does not significantly increase the

probability of an accident or the

consequences of an accident previously

identified. The support systems for

minimizing the consequences of transients

and accidents are not affected by the

proposed amendment.

New Fuel Design

The use of reload quantities of ATRIUM–

9B fuel at Dresden does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated in the [Final Safety Analysis

Report] FSAR. The ATRIUM–9B fuel is

generically approved for use as a reload BWR

fuel type (Reference: ANF–89–014(P)(A)

Revision 1 Supplement 1, Generic

Mechanical Design for Advanced Nuclear

Fuels 9X9–IX and 9X9–9X BWR Reload

Fuel). Limiting postulated occurrences and

normal operation have been analyzed using

NRC-approved methods for the ATRIUM–9B

fuel design to ensure that safety limits are

protected and that acceptable transient and

accident performance is maintained.

The reload fuel has no adverse impact on

the performance of in-core neutron flux

instrumentation or CRD response. The

ATRIUM–9B fuel design will not adversely

affect performance of neutron

instrumentation nor will it adversely affect

the movement of control blades relative to

the current Dresden fuel type, the Siemens

manufactured 9x9–2. The exterior

dimensions of the ATRIUM–9B fuel have

been evaluated by ComEd; the ATRIUM–9B

fuel design provides adequate clearances

relative to the co-resident 9x9–2 fuel. Thus,

no increased interactions with the adjacent

control blade or nuclear instrumentation are

created. Additionally, given the above

mentioned overall envelope similarities, no

problems are anticipated with other station

equipment such as the fuel storage racks, the

new fuel inspection stand and the spent fuel

storage pool fuel preparation machine.

Therefore, the probability of adverse

interactions between the ATRIUM–9B fuel

and components in the core and fuel

handling equipment is not significantly

increased.

The ATRIUM–9B design is neutronically

compatible with the existing fuel types and

core components in the Dresden core. SPC

tests have demonstrated that the ATRIUM–

9B fuel design is hydraulically compatible

with the co-resident 9x9–2 fuel. The bundle

pressure drop characteristics of the ATRIUM
17228 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

9B bundle are similar to those of the 9x9–2

fuel design, hence core thermal-hydraulic

stability characteristics are not adversely

affected by the ATRIUM–9B design. Cycle

stability calculations are performed by SPC.

Therefore, the probability of thermal

hydraulic instability is not significantly

increased.

Evaluations of the Dresden Emergency

Procedures and UFSAR Chapter 15 AOOs are

being performed to ensure that the use of the

ATRIUM–9B fuel at Dresden does not alter

any assumptions previously made in

evaluating the radiological consequences of

an accident at Dresden Units 2 and 3.

Therefore, the radiological consequences of

accidents are not significantly increased.

Methods approved by the NRC are being

used in the evaluation of fuel performance

during normal and abnormal operating

conditions. The ComEd and SPC methods to

be used for the cycle specific transient

analyses have been previously NRC

approved. The proposed methodologies are

administrative in nature and do not

significantly affect any accident precursors or

accident results; as such, the proposed

change to the listing of the SPC

methodologies for Dresden does not

significantly increase the probability or

consequences of any previously evaluated

accidents.

The description of the fuel is modified to

include the water box design of the NRC

approved ATRIUM–9B fuel type.

Review of the above concludes that the

probability of occurrence and the

consequences of an accident previously

evaluated in the safety analysis report have

not been significantly increased.

* * * * *

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated:

Creation of the possibility of a new or

different kind of accident would require the

creation of one or more new precursors of

that accident. New accident precursors may

be created by modifications of the plant

configuration, including changes in

allowable modes of operation.

Licensing Methods and Models

The proposed Technical Specification

amendment reflects previously approved SPC

methodology used to analyze normal

operations, including AOOs, and to

determine the potential consequences of

accidents. In accordance with FSAR Chapter

15, the same accidents and transients will be

analyzed with the new fuel and method as

have been previously performed. As stated

above, the proposed changes do not permit

modes of reactor operation which differ from

those currently permitted; therefore, the

possibility of a new or different kind of

accident is not created. Plant support

equipment is not affected by the proposed

changes; therefore, no new failure modes are

created.

New Fuel Design

The basic design concept of a 9x9 fuel pin

array with an internal water box has been

used in various lead assembly programs and

in reload quantities in Europe since 1986.

WNP–2 has loaded reload quantities since

1991. Eight lead ATRIUM–9B assemblies

were loaded into Dresden 2 during Cycle 15.

Approximately 650 water box assemblies

have been irradiated in the United States

through 1995, with a substantially higher

number being irradiated overseas. The NRC

has reviewed and approved the ATRIUM–9B

fuel design (Reference: ANF–89–014(P)(A)

Revision 1 Supplement 1, Generic

Mechanical Design for Advanced Nuclear

Fuels 9X9–IX and 9X9–9X BWR Reload

Fuel). The similarities in fuel design and

operation between the ATRIUM–9B and the

9x9–2, and the previous Boiling Water

Reactor experience with Siemens fuel,

indicate there would be no new or different

types of accidents for Dresden than have

been considered for the existing fuel.

Therefore, the use of ATRIUM–9B fuel at

Dresden does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

* * * * *

3. Involve a significant reduction in the

margin of safety for the following reasons:

The existing margin to safety is provided

by the existing acceptance criteria (e.g., 10

CFR 50.46 limits). The proposed Technical

Specification amendment reflects previously

approved SPC methodology used to

demonstrate that the existing acceptance

criteria are satisfied. The revised LOCA

methodology has been previously reviewed

and approved by the USNRC for application

to reload cores of BWRs. References for the

Licensing Topical Reports which document

this methodology, and include the Safety

Evaluation Reports prepared by the USNRC,

are added to the Reference section of the

Technical Specifications as part of this

amendment.

Licensing Methods and Models

The proposed amendment does not involve

changes to the existing operability criteria.

NRC approved methods and established

limits (implemented in the COLR) ensure

acceptable margin is maintained. The ComEd

and SPC reload methodologies for the

ATRIUM–9B reload design are consistent

with the Technical Specification Bases. The

Limiting Conditions for Operation are taken

into consideration while performing the

cycle specific and generic reload safety

analyses. USNRC approved methods are

listed in Specification 6.9.A of the Technical

Specifications.

Analyses performed with USNRCapproved methodology have demonstrated

that fuel design and licensing criteria will be

met during normal and abnormal operating

conditions. The same margins of safety will

continue to be utilized by SPC (e.g., limits on

peak cladding temperature, cladding

oxidation, plastic strain). Therefore, there is

not a significant reduction in the margin of

safety.

New Fuel Design

The exterior dimensions of the ATRIUM–

9B fuel assembly result in equivalent

clearances relative to the co-resident 9x9–2

fuel. Thus, no increased interactions with the

adjacent control blade and nuclear

instrumentation are created. The change does

not adversely impact equipment important to

safety; therefore the margin of safety is not

significantly reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Robert A. Capra.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of amendment request: March

18, 1997.

Description of amendment request:

The proposed amendments would

change the Technical Specifications

(TS) by increasing the High Pressure

Coolant Injection (HPCI) isolation

setpoint from greater than/equal to 80

psig to greater than/equal to 100 psig.

The licensee has requested the change

to ensure consistency between the

Updated Final Safety Analysis Report

(UFSAR), design basis documents and

the TS. The function of the setpoint is

to assure the HPCI turbine steam supply

is isolated in the event that the reactor

scram supply pressure falls below the

stall pressure of the HPCI turbine and

the system seals are no longer effective

in controlling the release of potentially

contaminated steam.

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 Low Reactor Pressure isolation of the

HPCI steam supply lines is provided to

prevent damage to the HPCI turbine when the

reactor steam pressure has decreased below

that required to provide adequate motive

force to operate the system. The steam supply

isolation low reactor pressure setpoint is not

an assumed initiator or contributor to any

previously evaluated accident and therefore

this change does not involve an increase in

the probability of an accident previously

evaluated at Dresden Station.

The Lower Reactor Pressure isolation of the

HPCI steam supply lines is described in the
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17229

plant safety analysis as a backup protection

to other system and facility design features

which provide assurance that accident

transients will not result in failures of the

system which contribute significantly to the

consequences of the initiating accident. The

low reactor pressure isolation signal provides

backup to other isolation signals to ensure

isolation will occur, minimizing the radiation

dose as a result of steam leakage past the

turbine seals in the event of a locked rotor

due to damage from liquid carryover due to

postulated swell in the reactor vessel.

These analyses assume the isolation

function occurs at 100 psig, and the proposed

setpoint of greater than or equal to 100 psig

is consistent and conservative with respect to

these assumptions. Because the isolation

function is not an accident initiator and the

revised setpoint ensures that the isolation

function continues to minimize radiological

consequences, the consequences of any

accident previously evaluated is not

increased by the proposed changes.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed change administratively

increases the Low Reactor Vessel Pressure

trip setpoint which initiates HPCI isolation.

This change does not result in any new or

different modes of operation. The proposed

change increases the setpoint at which the

HPCI turbine steam supply will be isolated

as the reactor vessel pressure decreases

following a postulated accident. The

proposed new setpoint is conservative with

respect to the existing TS limit, i.e. the new

limit of greater than or equal to 100 psig is

consistent and permitted by the existing limit

of greater than or equal to 80 psig. The

change assures that the Trip Setpoint in the

TS accurately reflects the design basis and

UFSAR described limits.

Because the proposed change does not

result in any new modes of plant operation

and administratively increases the system

isolation setpoint in a conservative manner,

the proposed change does not create the

possibility of a new or different kind of

accident from those previously evaluated.

3. Involve a significant reduction in the

margin of safety because:

The Trip Setpoint provides assurance that

the HPCI turbine cannot be operated with a

steam supply pressure too low to drive the

turbine and pump. The isolation assures that

the turbine does not stall and minimizes the

potential for the release of radioactivity

which results from steam leakage past the

turbine seals. The proposed change increases

the setpoint, ensuring that the required

isolation occurs at a higher pressure which

is more conservative, i.e. it assures the

turbine is isolated before the inlet steam

pressure falls to the stall pressure of the HPCI

turbine and leakage occurs. The greater than

or equal to 100 psig limit is well below the

range of reactor vessel pressure for which

HPCI is required to perform its safety

function. Therefore, the margin of safety

provided by the function of the HPCI

isolation on low reactor vessel pressure is

increased by the proposed TS change, and

this change will not involve a reduction in

the margin of safety.

As described, the proposed amendment for

Dresden will not reduce the availability of

systems required to mitigate accident

conditions. Neither are new or significantly

different modes of operation proposed.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

Guidance has been provided in ‘‘Final

Procedures and Standards on No Significant

Hazards Considerations,’’ Final Rule, 51 FR

7744, for the application of standards to

license change requests for determination of

the existence of significant hazards

considerations. This document provides

examples of amendments which are and are

not considered likely to involve significant

hazards considerations.

This proposed amendment does not

involve any irreversible changes, a significant

relaxation of the criteria used to establish

safety limits, a significant relaxation of the

bases for the limiting safety system settings

or a significant relaxation of the bases for the

limiting conditions for operations. Therefore,

based on the guidance provided in the

Federal Register and the criteria established

in 10 CFR 50.92(c), the proposed change does

not constitute a significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request:

December 27, 1995, as supplemented

September 4, October 18, and November

26, 1996.

Description of amendment request:

The proposed amendment would revise

technical specifications (TS) related to

electrical power systems. The proposed

changes include revisions to limiting

conditions for operation (LCO), LCO

applicability and action statements,

allowed outage times (AOT),

surveillance requirements (SR), and

administrative controls. The changes

add new requirements, revise or delete

existing requirements, relocate certain

existing requirements to other licensee

controlled documents, and editorially

restructure the proposed requirements

to closely emulate the electrical power

system requirements of NUREG–1432,

‘‘Standard Technical Specifications for

Combustion Engineering Plants,’’ (STS).

The proposed requirements differ from

the requirements of the STS where

necessary to reflect features unique to

the Palisades design. Each proposed

change has been classified by the

licensee as Administrative, Relocated,

More Restrictive, or Less Restrictive.

Basis for proposed no significant

hazards consideration determination: 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; (2) create

the possibility of a new or different kind

of accident from any previously

evaluated; or (3) involve a significant

reduction in a margin of safety. 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:

Evaluation of ADMINISTRATIVE,

RELOCATED, and MORE RESTRICTIVE

changes:

ADMINISTRATIVE changes and

RELOCATED changes move requirements,

either within the TS or to documents

controlled under 10 CFR 50.59, or [clarify]

existing TS requirements, without affecting

their technical content. Since

ADMINISTRATIVE and RELOCATED

changes do not alter the technical content of

any requirements, they cannot involve a

significant increase in the probability or

consequences of an accident previously

evaluated, create the possibility of a new or

different kind of accident from any

previously evaluated, or involve a significant

reduction in a margin of safety.

MORE RESTRICTIVE changes only add

new requirements, or revise existing

requirements to result in additional

operational restrictions. Since the TS, with

all MORE RESTRICTIVE changes

incorporated, will still contain all of the

requirements which existed prior to the

changes; MORE RESTRICTIVE changes

cannot involve a significant increase in the

probability or consequences of an accident

previously evaluated, create the possibility of

a new or different kind of accident from any

previously evaluated, or involve a significant

reduction in a margin of safety.

Evaluation of LESS RESTRICTIVE changes:

1. Do these LESS RESTRICTIVE changes

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Change 3 revised the requirement for

operable AC sources, using more general

wording than the existing TS. The existing

LCO requires that two explicitly specified

transformers be operable; the proposed LCO

requires that two qualified offsite circuits be

operable. The proposed LCO will allow
17230 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

substitution of Safeguards Transformer 1–1

for Station Power Transformer 1–2 as a

required AC source, but the quantity and

quality of required offsite AC sources is

unaffected. Since the capability and

qualification of Safeguards Transformer 1–1

are equivalent to those of the Station Power

transformer, neither the probability or

consequences of an accident previously

evaluated will be increased.

Change 10 is less restrictive only in its

allowance of a 72 hour AOT for an

inoperable offsite source instead of the 24

hour AOT currently required. The change

also makes a considerably more restrictive

change by eliminating the allowance, based

on submittal of a report, for continuous

operation with Startup Transformer 1–2

inoperable. Changing an AOT, alone, cannot

increase the probability or consequences of

an accident previously evaluated.

Change 14 allows, for an inoperable DG

[diesel generator], verification that no

common cause failure is involved in lieu of

test starting the other DG. The intent of the

test starting requirement is to verify that

there is no common cause failure which also

makes the other DG inoperable. The

proposed action statement thereby

accomplishes the same objective as that it

replaces. Since the proposed action statement

accomplishes the same objective as the one

it replaces, operation in accordance with the

proposed change will not increase the

probability or consequences of an accident

previously evaluated.

Change 21 revises the SR for the DG

starting test. [The ‘‘Less Restrictive’’ elements

of the change eliminate the requirement to

vary use of the A and B starting circuits for

each monthly test, because the DG is not

assumed to be single failure proof; and

eliminate requirements that the DGs be

manually started and that they be

synchronized from the control room, because

no practical alternatives exist for

accomplishing these actions]. The proposed

change does not alter any plant operating

conditions, operating practices, equipment

settings, or equipment capabilities.

Therefore, operation of the facility in

accordance with the proposed change will

not involve an increase in the probability of

an accident. Change 21 requires more

rigorous testing of the DGs than required by

the existing Technical Specifications. The

more rigorous testing is intended to provide

additional assurance that the DGs are capable

of performing their design function and

should, therefore, involve a reduction, rather

than an increase, in the consequences of

those accidents previously evaluated.

Change 25 revises the SR for testing the

fuel transfer system. The proposed change

does not alter any plant operating conditions,

operating practices, equipment settings, or

equipment capabilities. Therefore, operation

of the facility in accordance with the

proposed change will not involve an increase

in the probability of an accident. The only

‘‘Less Restrictive’’ feature of proposed SR is

test interval extension from ‘‘each month’’ to

‘‘each 92 days.’’ Changing a surveillance

frequency, alone, cannot increase the

probability or consequences of an accident

previously evaluated.

Change 26 revises the station battery SRs.

The proposed monthly and quarterly battery

SRs contain all of the test requirements of the

existing SRs with two exceptions: (1) The

proposed interval for measuring each cell

voltage is ‘‘each 92 days’’ instead of the

existing ‘‘every month’’ and (2) the

requirement to record the amount of water

added has been deleted. Changing a

surveillance frequency or deleting a

maintenance record cannot increase the

probability or consequences of an accident

previously evaluated.

2. Do changes create the possibility of a

new or different kind of accident from any

previously evaluated?

Change 3 only involves the specified offsite

power sources. Since the Loss of Offsite

Power is already considered in the accident

analyses, operating the facility in accordance

with Change 3 will not create the possibility

of a new or different kind of accident from

any previously evaluated.

Change 10 revises an AOT; Change 14

revises a required action; Change 21 revises

a testing requirement; Changes 25 and 26

revise a surveillance interval; and Change 26

deletes the requirement for a maintenance

record. None of these proposed changes alter

any plant operating conditions, operating

practices, equipment settings, or equipment

capabilities. Therefore, operation of the

facility in accordance with the proposed

changes will not create the possibility of a

new or different kind of accident from any

previously evaluated.

3. Do changes involve a significant

reduction in a margin of safety?

Change 3 does not alter the quantity or

quality of offsite sources required to be

available. Therefore, operating the facility in

accordance with the proposed change will

not involve a reduction in a margin of safety.

Change 10 revises an AOT; Change 14

revises a required action, Change 21 revises

a testing requirement; Changes 25 and 26

revise a surveillance interval; and Change 26

deletes the requirement for a maintenance

record. These proposed changes do not alter

any plant operating conditions, operating

practices, equipment settings, or equipment

capabilities. Therefore, operating the facility

in accordance with the proposed change will

not involve a reduction in a margin of safety.

The licensee’s September 4, 1996,

supplement stated that three of the

proposed changes contained in the

supplement were not addressed in the

December 27, 1995, no significant

hazards analysis. The changes involved

TS requirements that would be deleted.

Equivalent requirements would be

incorporated in the FSAR or other

documents subject to the controls of 10

CFR 50.59. The licensee’s analysis of the

issue of no significant hazards

consideration for these changes is

presented below:

1. Do changes which relocate a

requirement from the TS to documents which

are controlled under 10 CFR 50.59 involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

10 CFR 50.59 specifically prohibits

[without obtaining prior NRC review and

approval] changes to the facility as described

in the safety analysis report, and to

procedures described in the safety analysis

report ‘‘if the probability of occurrence or the

consequences of an accident or malfunction

of equipment important to safety previously

evaluated in the safety analysis report may be

increased’’. Since the conditions which limit

changes performed under 50.59 are more

restrictive than the conditions which define

changes considered to involve a significant

hazards consideration, relocation of a

requirement from the TS to the FSAR [Final

Safety Analysis Report] or to documents

which are referenced by the FSAR cannot

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Do changes which relocate a

requirement from the TS to documents which

are controlled under 10 CFR 50.59 create the

possibility of a new or different kind of

accident from any previously evaluated?

10 CFR 50.59 specifically prohibits

[without obtaining prior NRC review and

approval] changes to the facility as described

in the safety analysis report, and to

procedures described in the safety analysis

report ‘‘if a possibility for an accident or

malfunction of a different type than any

evaluated previously in the safety analysis

report may be created’’. Since the conditions

which limit changes performed under 50.59

are more restrictive than the conditions

which define changes considered to involve

a significant hazards consideration,

relocation of a requirement from the TS to

the FSAR or to documents which are

referenced by the FSAR cannot create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Do these changes which relocate a

requirement from the TS to documents which

are controlled under 10 CFR 50.59 involve a

significant reduction in a margin of safety?

10 CFR 50.59 specifically prohibits

[without obtaining prior NRC review and

approval] changes to the facility as described

in the safety analysis report, and to

procedures described in the safety analysis

report ‘‘if the margin of safety as defined in

the basis for any technical specification is

reduced’’. Since the conditions which limit

changes performed under 50.59 are more

restrictive than the conditions which define

changes considered to involve a significant

hazards consideration, relocation of a

requirement from the TS to the FSAR or to

documents which are referenced by the

FSAR cannot involve a significant reduction

in a margin of safety.

The NRC staff has reviewed the

licensee’s analyses 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17231

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: John N.

Hannon.

Duquesne Light Company, et al., Docket

Nos. 50–334 and 50–412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: March

10, 1997.

Description of amendment request:

The proposed amendment would

modify Unit 1 Technical Specification

(TS) 5.2.1 to add ZIRLO as fuel

assembly material and add reference to

Nuclear Regulatory Commission

approved Topical Report, WCAP–12610,

‘‘Vantage+ Fuel Assembly Reference

Core Report’’, to TS 6.9.1.12 for both

units.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The methodologies used in the accident

analyses have been modified to reflect the

requirements provided in WCAP–12610,

VANTAGE+ Fuel Assembly Reference Core

Report. Reference to this NRC approved

ZIRLO topical report has been added to

Specification 6.9.1.12, for both units to

ensure the analytical methods used to

determine the core operating limits are

consistent with those previously approved by

the NRC. The proposed changes do not

change or alter the design assumptions for

the systems or components used to mitigate

the consequences of an accident. Use of

ZIRLO fuel rod material does not adversely

affect fuel performance or impact nuclear

design methodology. Therefore, accident

analysis results are not impacted.

The operating limits will not be changed

and the analysis methods to demonstrate

operation within the limits will remain in

accordance with NRC approved

methodologies. Other than the changes to the

fuel assemblies, there are no physical

changes to the plant associated with this

technical specification change. A safety

analysis will continue to be performed for

each cycle to demonstrate compliance with

all fuel safety design bases.

VANTAGE 5H fuel assemblies with ZIRLO

fuel rods meet the same fuel assembly and

fuel rod design bases as other VANTAGE 5H

fuel assemblies. In addition, the 10 CFR

50.46 criteria are applied to the ZIRLO fuel

rods. The use of these fuel assemblies will

not result in a change to the reload design

and safety analysis limits. Since the original

design criteria are met, the ZIRLO fuel rods

will not be an initiator for any new accident.

The fuel rod material is similar in chemical

composition and has similar physical and

mechanical properties as Zircaloy-4. Thus,

the fuel rod integrity is maintained and the

structural integrity of the fuel assembly is not

affected. ZIRLO improves corrosion

performance and dimensional stability. No

concerns have been identified with respect to

the use of an assembly containing a

combination of Zircaloy-4 and ZIRLO fuel

rods.

The dose predictions in the safety analyses

are not sensitive to the fuel rod material

used; therefore, the radiological

consequences of accidents previously

evaluated in the safety analysis remain valid.

A reload analysis is completed for each cycle,

in accordance with NRC approved

methodologies. Therefore, the proposed

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

VANTAGE 5H fuel assemblies with ZIRLO

fuel rods satisfy the same design bases as

those used for other VANTAGE 5H fuel

assemblies. All design and performance

criteria continue to be met and no new

failure mechanisms have been identified. The

ZIRLO fuel rod material offers improved

corrosion resistance and structural integrity.

The proposed changes do not affect the

design or operation of any system or

component in the plant. The safety functions

of the related structures, systems, or

components are not changed in any manner,

nor is the reliability of any structure, system,

or component reduced. The changes do not

affect the manner by which the facility is

operated and do not change any facility

design feature, structure, or system. No new

or different type of equipment will be

installed. Since there is no change to the

facility or operating procedures, and the

safety functions and reliability of structures,

systems, or components are not affected, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The use of Zircaloy-4, ZIRLO, or stainless

steel filler rods in fuel assemblies will not

involve a significant reduction in the margin

of safety because analyses using NRC

approved methodology will be performed for

each configuration to demonstrate continued

operation within the limits that assure

acceptable plant response to accidents and

transients. These analyses will be performed

using NRC approved methods that have been

approved for application to the fuel

configuration.

Use of ZIRLO as fuel rod material does not

change the VANTAGE 5H reload design and

safety analysis limits. The use of these fuel

assemblies will take into consideration the

normal core operating conditions allowed in

the technical specifications. For each reload

core, the fuel assemblies will be evaluated

using NRC approved reload design methods,

including consideration of the core physics

analysis peaking factors and core average

linear heat rate effects.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a significant reduction in margin

with respect to plant safety as defined in the

UFSAR [Updated Final Safety Analysis

Report] or any plant technical specification

BASES.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 17,

1996.

Description of amendment request:

The proposed amendment would reflect

that the name of Louisiana Power &

Light Company, which is licensed to

own and possess Waterford 3, has been

changed to Entergy Louisiana, Inc.

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

The proposed change documents changing

the legal name of the company. The proposed

change will not affect any other obligations.

The company will still own all of the same

assets, they still serve the same customers,

and all existing obligations and commitments

will continue to be honored. 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

The administrative changes in the

Operating License requirements do not

involve any change in the design of the plant.

Therefore, the proposed change will not

create the possibility of a new or different
17232 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

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

The proposed change is administrative in

nature and does not reduce the level of safety

imposed by any current requirement.

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: October

16, 1996.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) action

requirements 3.2.1 and 3.2.4 and their

associated surveillance requirements to

extend the allowable time for the Core

Operation Limit Supervisory System

(COLSS) to be out of service by

monitoring for adverse trends in the

linear heat rate (LHR) and departure

from nucleate boiling (DNBR) limits.

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

The proposed change does not modify the

requirement to operate within the alternate

LHR and DNBR limits nor does it modify the

actual LHR or DNBR limits themselves. In the

case of exceeding a COLSS calculated [power

operating limit] POL, Entergy agrees that

corrective action should be initiated

promptly to bring the LHR and DNBR within

their respective limits and, in this case, a 15

minute time limit is appropriate. However, in

the case of exceeding a [core protection

calculator] CPC calculated operating limit

following the loss of COLSS, it is clear that

simply because COLSS execution was lost

does not mean that the plant is operating

outside the range of conditions assumed in

the Chapter 15 Safety Analysis and, in this

case, a 15 minute time limit is not

appropriate. An increase from 2 hours to 8

hours to regain the monitoring capabilities of

COLSS would not significantly increase the

probability of exceeding the actual LHR or

DNBR power operating limits since the

increase in COLSS out-of-service time will be

compensated for by monitoring for adverse

trends of the important CPC calculated

parameters (DNBR Margin and LHR). Further,

since the proposed change will result in

maintaining steady-state conditions while

monitoring for adverse trends, it will be

easier for the operators to detect any

abnormal occurrence that has the potential to

degrade either the LHR or the DNBR.

The primary consideration in extending

the COLSS out of service time limit is the

remote possibility of a slow, undetectable

transient that degrades the LHR and/or DNBR

slowly over the 8 hour period and is then

followed by an [anticipated operational

occurrence] AOO or an accident. The

parameters normally monitored by COLSS

which have the potential for degrading the

LHR and DNBR if no corrective action is

taken are: Reactor Coolant System (RCS) flow

rate, axial and radial power distributions,

core inlet temperature, core power, RCS

pressure and azimuthal tilt. Of these

parameters, core inlet temperature, core

power, and RCS pressure are easily

monitored by the plant operators using

various safety-grade, Redundant Control

Room indications and, therefore, changes in

these parameters are readily apparent.

Further, operating experience at Waterford 3

and other [Combustion Engineering] CE

nuclear steam supply systems using the same

reactor coolant pumps (RCPs) as Waterford

has shown that measurable changes in RCP

>Ps (which COLSS uses to calculate RCS

flow) are very rare and when they do occur,

involve abrupt step changes in flow which

are readily apparent; hence, the probability of

a slow degradation in the RCS flow rate is

exceedingly small. Thus, the parameters that

comparatively (although still remote) pose

the most potential for a degradation in the

core thermal margin when COLSS is out of

service relate to the axial and radial core

power distributions and the azimuthal tilt.

These parameters are discussed below.

Axial xenon oscillations are a normal

consequence of the Waterford 3 core design,

particularly near the end of core life. As a

result, Waterford 3 operations personnel are

instructed, per operating procedure OP–10–

001, General Plant Operations, to maintain

strict control over the axial power shape in

the core. Although the primary reason for

axial shape control is to maintain an even

fuel burnup throughout the core, it also

results in maintaining the axial power shapes

well within the limits assumed in the safety

analysis. Typically, axial shape control

practiced at Waterford 3 maintains the axial

shape index (ASI) within 0.05 ASI units of

the equilibrium shape index (ESI), which is

normally very near 0.0.

Hypothetically, the most severe situation

which could be postulated to occur, although

again remote, would be if COLSS execution

was lost just when the plant operators were

ready to take manual action to return the ASI

value to within the ESI + 0.05 control band.

Since a full xenon oscillation takes

approximately 26 hours, there would be

about 6 hours from the time that control

action would normally be taken to the time

that the ASI reached its peak value (i.e., it

takes one quarter cycle for the ASI to travel

from its ESI value to its peak value). Since

abnormal operating procedure OP–901–501,

PMC or Core Operating Limit Supervisory

System Inoperable, will be revised to require

the CPC calculated LHR and DNBR trends to

be monitored every 15 minutes (see below),

any significant change in the axial shape

index will be apparent through a change in

these CPC calculated values. Hence, due to

the attention given the axial power

distribution, both when COLSS is in service

as well as when COLSS is out of service it

is very improbable that a change in ASI

during eight hours of steady-state operation

with COLSS out of service could be either

undetected or lead to a condition that placed

the reactor outside the range of initial

conditions that were assumed in the safety

analysis.

With regards to azimuthal tilt, there is very

rarely any significant change in this

parameter as long as all [control element

assembly] CEAs are properly aligned. The

only real contributor to a rapid increase in

azimuthal tilt would be an inadvertent CEA

drop; however, since the probability of a CEA

drop is very low, the likelihood of this event

occurring within the eight hour time limit is

even lower. In the unlikely event that a CEA

drop did occur, the Control Element

Assembly Calculators (CEACs) provide a

safety-grade, redundant means of alerting the

operators that corrective action is necessary.

Thus, the potential for a degradation in

azimuthal tilt during eight hours of steadystate operation following the loss of COLSS

is both highly unlikely and relatively easy to

detect using instrumentation already

available in the Control Room.

As previously stated, upon approval of the

proposed change plant personnel will revise

abnormal operating OP–901–501, PMC or

Core Operating Limit Supervisory System

Inoperable, to monitor for adverse trends of

the CPC calculated values of LHR and DNBR.

Currently, this procedure requires that the

monitoring frequency for LHR and DNBR be

increased to once every 15 minutes on a loss

of COLSS.

Extending the time to restore the CPC

calculated LHR and DNBR to within the

acceptable operating range from 2 hours to 8

hours is being proposed to assure that COLSS

can be restored thus decreasing the

probability of an avoidable challenge to the

reactor protection system (RPS) during a

power reduction. It is possible that the

required power reductions may exceed 25%

near the end of the fuel cycle. These large

power reductions result in a rapid increase

in xenon concentration, changes in ASI, and

a subsequent decrease in cold leg

temperature (T-cold) that may be difficult to

control. Accordingly, given the potential for
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17233

power reductions of this magnitude, it is

appropriate to extend the time allowed to

restore COLSS so that a power reduction may

be unnecessary.

Taken in total, the proposed changes will

reduce the number of potentially

unnecessary power reductions by allowing

more time for COLSS to be restored along

with the advantages of trend monitoring in

detecting an adverse trend expeditiously. The

proposed change will result in significant

operational benefits while continuing to

maintain a high degree of confidence that the

core conditions remain well within the range

of values assumed in the safety 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

The proposed change does not alter the

current power operating limits nor does it

involve any changes to COLSS or CPC

software. There has been no physical change

to plant systems, structures or components

nor will the proposed change affect the

ability of any of the safety-related equipment

required to mitigate AOOs or accidents. The

only significant change associated with the

proposed amendment involves changes to the

operating procedures used when COLSS is

out-of-service. All revisions to operating

procedures will be reviewed and approved

by appropriate plant personnel as required by

the Administrative Controls (Section 6) in the

Waterford 3 Technical Specifications.

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

The intent of [limiting conditions for

operation] LCOs 3.2.1 and 3.2.4 is to

maintain the reactor within the range of

initial conditions that was assumed in the

Safety Analysis. Maintaining the LHR within

the specified range ensures that in the event

of a LOCA, the fuel cladding temperature

will not exceed the 2200°F limit imposed by

10CFR46 [10 CFR Part 46]. Maintaining the

DNBR within the specified range ensures that

no AOO will result in a violation of the

[Specified Acceptable Fuel Design Limits]

SAFDLs and that no postulated accident will

result in consequences more severe than

those described in Chapter 15 of the [Final

Safety Analysis Report] FSAR. Since there

has been no change to the requirement to

operate the reactor within the LHR and

DNBR limits and no change to the actual LHR

and DNBR limits themselves, the accident

analyses described in Chapter 15 of the FSAR

will not be affected and will therefore remain

bounding.

The proposed change will reduce the

number of potentially unnecessary power

reductions along with the rate at which the

power reductions are accomplished.

Maintaining steady-state conditions for up to

eight hours after the loss of COLSS while

monitoring the CPC LHR/DNBR for trends,

provides plant personnel with a reasonable

period of time to return COLSS to service

while continuing to maintain a high degree

of confidence that the core conditions remain

well within the range of values assumed in

the safety analysis. In fact, monitoring for

trends in LHR and DNBR Margin increases

the margin of safety by allowing the

anticipation of degradation in LHR or DNBR

Margin. Moreover, by reducing the number of

plant transients there will be an attendant

reduction in probability of an AOO and

subsequent RPS actuation. 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: October

16, 1996.

Description of amendment request:

The following changes to the Waterford

Steam Electric Station, Unit 3,

Technical Specifications are proposed:

1) Relocation of certain administrative

controls to the Quality Assurance

Program Manual (QAPM) as described

in Nuclear Regulatory Commission

Administrative Letter 95–06,

‘‘Relocation of Technical Administrative

Controls related to Quality Assurance’’;

2) Change shift coverage from 8-hour

day, 40-hour weeks to an option of 8 or

12 hour days and nominal 40-hour

weeks; 3) Make certain editorial changes

to the titles of certain organizational

positions.

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

The conditions as they exist in the present

Technical Specifications do not have an

affect on either the probability or

consequences of a previously evaluated

accident. These changes also will have no

impact to increase either the probability or

consequences of a previously evaluated

accident.

The proposed changes will have no affect

on design basis accidents nor will the change

directly affect any material condition of the

plant that could directly contribute to

causing or mitigating the effects of an

accident.

Therefore, the proposed changes 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

The proposed changes will not alter the

operation of the plant or the manner in

which it is operated. The changes do not

involve a design change and do not introduce

any new failure modes.

Therefore, the proposed changes 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 margin of

safety?

Response: No

The proposed changes are administrative

in nature and affect only Section 6.0 of the

Technical Specifications. The Waterford 3

margins of safety are defined in Sections 2

through 5 and are unaffected by these

changes. Moving the reviews from the TS to

the QAPM will have no affect on the margin

of safety because reviews will still be

performed. The only difference is the reviews

will be administratively controlled by the

QAPM. The QAPM is controlled by

10CFR50.54 so no changes can be made

which would lessen these commitments (i.e.,

remove or reduce the requirement for

procedure reviews) without prior NRC

approval.

Changing from an 8 hour to an 8 or 12 hour

shift will not have an adverse impact on

personnel performance. The NRC study

documented in NUREG CR–4248 has

identified that personnel errors have

decreased and productivity has increased

where this change has been implemented.

Therefore, the proposed changes 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
17234 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

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: March

27, 1997.

Description of amendment request:

The proposed change modifies

Technical Specification 3/4.5.2, ‘‘ECCS

Subsystems Modes 1, 2, and 3.’’ The

proposed change adds a surveillance

requirement to verify the Emergency

Core Cooling System (ECCS) piping is

full of water at least once per 31 days.

A change to the Technical Specification

Basis 3/4.5.2 and 3/4.5.3 has been

included to support this change.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. 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.

The proposed change will not affect the

assumptions, design parameters, or results of

any accident previously evaluated. The

proposed change does not add or modify any

existing equipment. The proposed change

adds a new surveillance requirement which

will minimize the likelihood of a pressure

transient occurring during system startup and

provide increased assurance that the ECCS

will perform its design basis function when

needed. The new [low pressure safety

injection] LPSI and [high pressure safety

injection] HPSI vent valves which may be

manipulated during this surveillance will be

administratively controlled and will be

locked close when not in use to prevent the

possibility of a flow diversion. This

surveillance requirement is consistent with

NUREG 1432.

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.

While new vent lines are being installed

under 10CFR50.59, this proposed change

adds only a new surveillance requirement to

Technical Specification 3/4.5.2 and therefore

does not involve modifications to any

existing equipment. The new vent valves,

when required, will be operated and

controlled in the same manner as existing

LPSI and HPSI vent valves. The new LPSI

and HPSI vent valves will be

administratively controlled and will be

locked close when not in use. This

surveillance requirement is consistent with

NUREG 1432.

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.

The functionality of ECCS is maintained

such that it is capable of performing its

design function as assumed in the Updated

Final Safety Analysis Report. Verifying the

ECCS is full of water at least once per 31 days

will minimize the likelihood of a pressure

transient occurring during system startup and

provide increased assurance that the ECCS

will perform its design basis function when

needed. This surveillance requirement is

consistent with NUREG 1432.

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: March

27, 1997.

Description of amendment request:

The proposed change modifies

Technical Specification (TS)

surveillance requirements 4.5.2.d.3 and

4.5.2.d.4. The proposed change specifies

granular trisodium phosphate

dodecahydrate (TSP), increases the

minimum required amount of TSP that

is maintained in containment during

power operation, and adjusts the TSP

sampling requirement accordingly. A

change to the TS Basis 3/4.5.2 has been

included to support this change.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. 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.

Granular trisodium phosphate

dodecahydrate is stored in the containment

lower level to raise the pH of the sump and

spray water following a LOCA. As the pH of

the water increases, more radioactive iodine

is kept in solution and the amount of

airborne radioactive leakage is decreased.

This also lessens the potential for boric acid

solution reacting with galvanized metal in

containment to release hydrogen. An

additional advantage of a higher pH is the

beneficial reduction in chloride stress

corrosion cracking of metal components in

the containment following an accident.

This chemical is an accident mitigator, not

an accident initiator in that it is not used

until after an accident has occurred. At the

time it goes into solution, the accident has

occurred, containment spray has been

activated and water has collected in the

sump. Therefore, increasing the Technical

Specification minimum amount verified to be

in containment or changing the sample

solution and sample size will not involve a

significant increase in the probability of an

accident previously evaluated.

At the time TSP goes into solution, the

accident has occurred, containment spray has

been activated and water has collected in the

containment sump. At Waterford 3, the

iodine partition factor is a constant 50% and

does not vary with pH as allowed in the

Standard Review Plan (SRP) revision 1. The

curve in SRP 6.5.2 revision 1 allows a

partition factor of at least 50% for

containment water at a pH of 6.5 or less. The

partition factor increases as pH rises. But, the

curve is based on sodium hydroxide which

is much more reactive than TSP. Therefore,

increasing the Technical Specification

minimum amount verified to be in the

containment, and corresponding sample size,

will not involve any significant increase in

the consequences probability of an accident

because no credit is taken for reducing the

amount of volatized iodine normally

associated with a 7.0 pH solution.

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.

The addition of more TSP does not

represent a significant change in the

configuration or operation of the plant.

Trisodium phosphate dodecahydrate is

currently present in the containment lower

level. Design Change 3491 which increases

the storage capacity of the TSP storage

baskets was evaluated in accordance with 10

CFR 50.59 and found not to involve an

unreviewed safety question.
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17235

Boric acid acts as a buffer to prevent the

pH from rising above approximately 8.1 as

TSP is dissolved. An internal study (EC–S96–

013 revision 0) has shown that given the

‘‘ratio of grams of TSP to liters of 3000 ppm

boron solution’’ stays less than 5.6, TSP

cannot increase pH above 8.2. As pH

increases, components composed of

aluminum, zinc, or copper become

vulnerable to corrosion. Branch Technical

Position MTEB 6–1 implies that a solution

pH greater than 7.5 enhances the chance for

hydrogen generation as a result of aluminum

corrosion. Waterford 3 administratively

limits the amount of aluminum in

containment to minimize the amount of

hydrogen expected during a DBA. Zinc is a

component of the paint applied to surfaces

inside containment. The hydrogen

recombiner design basis includes 464 square

feet (1040 pounds) of aluminum and 419,300

square feet (17,252 pounds) of metallic zinc.

Estimates of the amount of hydrogen

produced by the aluminum assumes that the

corrosive agent is sodium hydroxide—a

much more active chemical than is TSP.

Thus, the amount of hydrogen expected in

the FSAR for the hydrogen recombiner

bounds what would actually be produced by

TSP even at a pH of approximately 8.1.

The 4.5.2.d.3 proposed TSP to boron ratio

assures that pH cannot rise above 8.1 as long

as post accident in-containment boric acid

solution concentration is no greater than

3011 ppm boron and no less than 1504 ppm

boron. The main variable in post accident

concentration (the difference between 1504

and 3011) is the concentration in the RCS at

the time of the accident.

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.

Trisodium phosphate dodecahydrate is

stored in the containment lower level to raise

the pH of the sump and spray water

following a LOCA. As the pH of the water

increases, more radioactive iodine is kept in

solution and the amount of airborne

radioactive leakage is decreased. A neutral

pH also reduces the hydrogen generation

from the corrosion of the galvanized

materials in containment. An additional

advantage of a higher pH is the beneficial

reduction in chloride stress corrosion

cracking of metal components in the

containment following an accident.

Technical Specification 4.5.2.d.3 requires

verification that a minimum volume of TSP

is contained in the storage baskets in

containment. Nine previous runs of

surveillance requirement 4.5.2.d.4 (and

similar tests) showed that the TSP actually

used in the plant properly neutralized a

sample of water borated within RWSP boron

concentration limits. Boron concentrations of

eight of the sample solutions used in these

tests ranged from 1753 ppm to 2217 ppm and

resulted in a pH of 7.02 or greater. (The boron

concentration of one test performed in 1986

was unavailable.) The ratio 4 grams to 4 liters

is the amount of TSP needed to bring the

solution to a pH of at least 7.0 given that the

solution is in the 1753 to 2217 ppm Boron

range.

The amount of TSP in containment

currently is adequate assuming that RCS

boric acid concentration stays below 454

ppm. However, the fuel cycle is nearly over

and a restart with a refreshed core would

require substantially more boric acid. We

expect that the containment water would

reach approximately 2400 ppm under ideal

circumstances during cycle 9. During cycle

10, boron concentration in containment

could reach 3011 under those same ideal

conditions. As the maximum boron

concentration increases, there is a non-linear

increase in the amount of TSP needed to

raise solution pH to 7.0. Thus, we request

that the minimum amount of TSP in

containment required by 4.5.2.d.3 to be

increased from 97.5 cubic feet to 380 cubic

feet. This change also proposes to adjust the

4.5.2.d.4 specified increase that sample

solution and the TSP sample size

accordingly. This change will ensure the

safety injection containment sump, when

filled with water, will have an acceptable pH

following a LOCA. The test will not only

demonstrate that TSP is in the baskets but

also shows that the amount of TSP in

containment can neutralize the solution

expected in containment during any DBA.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety. The amount of iodine kept in solution

during a DBA is limited to 50%. Note, the pH

scale is logarithmic so that the amount of

TSP needed to raise pH to 7.0 is more than

three times the amount needed to reach 6.5.

Furthermore, the amount of hydrogen

generated during a DBA is over estimated by

the analysis when it used sodium hydroxide

as the corrosive agent rather than TSP.

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.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 17,

1996, as supplemented March 17, 1997.

Description of amendment request:

The proposed amendment would

modify Technical Specification (TS)

Section 3/4.4.5, Steam Generators, 3/

4.4.6, Reactor Coolant System Leakage,

and associated Bases to allow the

installation of tube sleeves as an

alternative to plugging to repair

defective steam generator tubes. The

proposed change would also specify the

Westinghouse topical reports to be used

for sleeve design and inspection, and

identify the inspection sample size for

repaired tubes. This application was

previously published in the Federal

Register on May 29, 1996, (61 FR

26938).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. Listing the specific Westinghouse

topical reports in the TS binds the South

Texas Project (STP) to the sleeve design and

inspection techniques identified in that

revision of the topical report. Any changes to

sleeve design or inspection technique would

require a separate TS amendment.

New TS Table 4.4–3, Steam Generator

Repaired Tube Inspection, identifies the

inspection sample size for steam generator

tubes that have already been repaired. This

table simply identifies inspection criteria and

associated actions for repaired tubes and

does not increase 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. Implementation of laser welded

sleeving maintains overall tube bundle

structural and leakage integrity conditions.

Providing specific Westinghouse topical

report references in the TS only serves to

identify which sleeve design and inspection

techniques are being employed at STP.

Likewise, the addition of Table 4.4–3 clarifies

the expected inspection samples for

previously repaired tubes. The addition of

Table 4.3–3 provides assurance that

previously repaired tubes will be inspected at

regular intervals and appropriate action taken

if the tube is found defective. Neither of these

additions to the TS will create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

Both of these changes are being added to

clarify the STP steam generator tube

inspection program and provide more

specific detail regarding steam generator tube

inspection samples and inspection

techniques. By requiring inspection of

previously repaired tubes, the margin of

safety is increased rather than decreased.
17236 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

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: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036–5869.

NRC Project Director: William D.

Beckner.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: January

28, 1997.

Description of amendment request:

The proposed amendment would

relocate the details of Technical

Specification (TS) Section 6.2.3 on the

Independent Safety Engineering Group

(ISEG) from the Administration Controls

section of the TSs and place these

details in the Updated Final Safety

Analysis Report (UFSAR) for South

Texas Project, Units 1 and 2. This

relocation is administrative only, and

would not render any changes to the

existing plant philosophy toward the

ISEG or any safety analysis. Section

6.2.3 would be deleted from the TSs and

removed from the table of contents for

Administrative Controls. Currently

UFSAR Section 13.4.2.2 describes the

ISEG, but not in the detail as the current

TSs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes move details from

the Technical Specifications [TSs] to the

Updated Final Safety Analysis Report

(UFSAR). The changes do not result in any

hardware or operating procedure changes.

The details being removed from the

Technical Specifications [TSs] are not

assumed to be an initiator of any analyzed

event. The UFSAR, which will contain the

removed Technical Specification [TS] details,

will be maintained using the provisions of 10

CFR 50.59 and is subject to the change

control process in the Administrative

Controls Section of the Technical

Specifications [TSs]. [In addition] any

changes to the UFSAR will be evaluated per

10 CFR 50.59, no increase in the probability

or consequences of an accident previously

evaluated will be allowed without prior NRC

[Nuclear Regulatory Commission] approval.

Therefore, the 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.

The proposed changes move details from

the technical Specifications [TSs] to the

Updated Final Safety Analysis Report

(UFSAR). The changes will not alter the plant

configuration (no new or different type of

equipment will be installed) or make changes

in methods governing plant operation. The

changes will not impose different

requirements, and adequate control of

information will be maintained. The changes

will not alter assumptions made in the safety

analysis and licensing basis. Therefore, the

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes move detail from

the Technical Specifications [TSs] to the

Updated Final Safety Analysis Report

(UFSAR). The changes do not reduce the

margin of safety since the relocation of

details [is an administrative action and] has

no impact on any safety analysis

assumptions. In addition, the detail

transposed from the Technical Specifications

[TSs] to the UFSAR are the same as the

existing Technical Specification [TS] [6.2.3].

[In addition] any future changes to the FSAR

will be evaluated per the requirements of 10

CFR 50.59, no reduction in a margin of safety

will be allowed without prior NRC approval.

[Therefore, the licensee concluded that the

changes 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 standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges, Learning Center,

911 Boling Highway, Wharton, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036–5869.

NRC Project Director: William D.

Beckner.

Northeast Nuclear Energy Company, et

al., Docket No. 50–245, Millstone

Nuclear Power Station, Unit No. 1, New

London, Connecticut

Date of amendment request: February

7, 1997.

Description of amendment request:

The proposed Technical Specification

changes would clarify and/or modify

instrument calibration, functional, and

response time requirements for

resistance temperature detector and

thermocouple testing. Also, certain

definitions would be clarified and/or

modified using applicable wording from

NRC’s NUREG–1433, ‘‘Standard

Technical Specifications,’’ Revision 1,

and industry recommendations.

Additionally, the change would relocate

the reactor protection system logic

response time value utilizing the

guidance provided by NRC’s Generic

Letter 93–08, ‘‘Relocation of Technical

Specification Tables of Instrument

Response Time Limits,’’ with the

exception of relocating the value to the

Technical Specifications Bases Section

instead of the Updated Final Safety

Analysis Report. The proposed

amendment is intended to clarify

instrumentation surveillance

requirements, thereby helping to ensure

proper testing of safety-related

components.

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:

Pursuant to 10 CFR 50.92, NNECO

[Northeast Nuclear Energy Company] has

reviewed the proposed changes and

concludes that the changes do not involve a

significant hazards consideration (SHC) since

the proposed changes satisf[y] the criteria in

10 CFR 50.92(c). That is, the proposed

changes do not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment continues to

ensure the surveillance requirements satisfy

the licensing basis. The current TS [technical

specifications] definition for Instrument

Functional Test requires injection of a

simulated signal into the primary sensor to

verify proper response. Current TS exempt

the sensors of specific instrument channels

where it is not practical to include them

within the functional test boundaries. Some

examples of these exemptions include

neutron monitoring system, turbine control

valve fast closure, and standby gas treatment

initiation radiation monitors. In these cases,

TS permit the performance of the functional

test by injection of a simulated electrical

signal into the measurement channel. The

proposed definition, which is consistent with

the STS [standard technical specifications]
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17237

definition, for CHANNEL FUNCTIONAL

TEST requires injection of the simulated

signal ‘‘as close to the sensor as practicable.’’

Therefore, the proposed definition is

consistent with the current TS definition and

its exemptions. The primary sensor is the

transmitter or switch or radiation monitor.

The definition does not include sensing

elements such as radiation detectors, flow

elements, acceleration relays or reference

legs.

This change will allow the channel

functional test to be performed by means of

any series of sequential, overlapping, or total

channel steps and aligns this methodology

with industry practice. This change does not

affect accident precursors and thus does not

involve a significant increase in the

probability of an accident previously

evaluated. The proposed change will allow a

simulated or actual signal to be used to

perform an Instrument or Channel Functional

Test. This change does not impose a

requirement to create an actual signal, nor

does it eliminate any restriction on

producing an actual signal. While creating an

‘‘actual’’ signal could increase the probability

of an event, existing procedures (and the 10

CFR 50.59 control of revisions to them)

dictate the acceptability of generating this

signal. The proposed change does not affect

the procedures governing plant operations or

the acceptability of creating these signals; it

simply would allow such a signal to be

utilized in evaluating the acceptance criteria

for the Instrument or Channel Functional

Test requirements. Therefore, the change

does not involve a significant increase in the

probability of an accident previously

evaluated. Because the method of initiation

will not affect the acceptance criteria of the

Instrument or Channel Functional Test, the

change does not involve a significant

increase in the consequences of an accident

previously evaluated.

Minor word differences from STS are

required to provide consistency with current

TS wording and support the current licensing

basis. These minor word differences

including Industry/TSTF [Technical

Specification Task Force] Standard Technical

Specification Change Traveler (TSTF–64) do

not alter the meaning of instrument testing in

the STS or change the current licensing basis.

Moving the RPS [Reactor Protection

System] Logic Response Time LCO [Limiting

Condition of Operation] description to the TS

definition section is an administrative change

and does not alter the original intent or

licensing basis.

Relocation of the RPS Logic Response Time

value from the TS to the Bases section

involves the use of an alternate regulatory

process for controlling the instrument

response time limit. The change does not

introduce any new modes of plant operation,

make any physical changes, alter any

operational setpoints, or change the

surveillance requirements. Any change in the

RPS logic response time value would be

evaluated pursuant to the requirements of 10

CFR 50.59.

The surveillance section editorial change

does not alter the meaning of surveillance

applicability. Providing RPS Logic Response

Time surveillance frequency and applicable

trip functions ensures proper testing of RPS

components and is consistent with industry

practice. An evaluation completed by GE

[General Electric] verified the applicable RPS

trip functions that require a specific logic

response time using the current accident

analysis as the basis. For trip functions where

no explicit credit is taken in the safety

analysis, the measurement of logic response

time is not important, and therefore, not

warranted. In addition, we have concluded,

that instrumentation response time

requirements (specified limits) other than

RPS logic are not important to test, especially

considering the long delays already

accounted for in the accident analyses

associated with the start of emergency power

sources, ECCS [Emergency Core Cooling

System] components, and containment

isolations, and that the non-RPS logic

response times, including response times of

other instrumentation such as radiation

monitors, are not part of the Millstone Unit

No. 1 licensing basis. The sensors associated

with all TS instrumentation are functionally

tested and calibrated to ensure proper

operation.

No physical change is being made to

instrument channels, or to any systems or

component that interfaces with the

instrumentation channels, therefore there is

no change in the probability or consequences

of any accident analyzed in the UFSAR

[Updated Final Safety Analysis Report].

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not result in

any design or physical configuration changes

to the instrumentation channels. Operation

incorporating the proposed change will not

impair the instrumentation channels from

performing as provided in the design basis.

Changing the TS to be consistent with

current industry practice adopted in STS will

help to prevent unnecessary removal and

potential damage of the temperature

detectors (for sensor calibration).

Clarification of RPS Logic Response Time

testing requirements consistent with the

current licensing basis will ensure proper

testing of safety-related components.

Wording changes to Instrument Calibration

and Functional Test definitions do not

involve a physical modification to the plant.

The injection of an actual or simulated signal

as close to the sensor as practical minimizes

the likelihood of any transients.

Minor word differences from STS are

required to provide consistency with current

TS wording and support the current licensing

basis. These minor word differences,

including Industry/TSTF Standard Technical

Specification Change Traveler (TSTF–64), do

not alter the meaning of instrument testing in

the STS or change the current licensing basis.

Moving the RPS Logic Response Time LCO

description to the TS definition section is an

administrative change and does not alter the

current licensing basis.

Relocation of the RPS Logic Response Time

value involves the use of an alternate process

for controlling the instrument response time

limits. Therefore, the above change does not

introduce any accident initiators as it does

not involve any new modes of plant

operation, make any physical changes, alter

any operational setpoints, or change the

surveillance requirements.

The surveillance section editorial change

does not alter the meaning of surveillance

applicability. Providing RPS Logic Response

Time surveillance frequency and applicable

trip functions ensures proper testing of RPS

components and is consistent with industry

practice.

Since the proposed changes in the

Technical Specifications do not adversely

impact the reliability of the RPS and other

automatic actuations, no new or different

kind of accident is created.

3. Involve a significant reduction in a

margin of safety.

Because the proposed change does not

involve the addition or modification of plant

equipment, is consistent with the existing

Technical Specifications, current industry

practices as outlined in NUREG 1433,

‘‘Standard Technical Specifications GE

Plants, BWR/4,’’ Revision 1, and with the

current design and licensing basis of the

Protective Instrumentation systems including

the accident analysis, no action will occur

that will involve a significant reduction in a

margin of safety.

The proposed change to allow the use of

an actual signal in addition to the existing

requirement, which limits use to a simulated

signal, will not affect functional test

acceptance criteria. Therefore, the proposed

change does not adversely affect the

reliability of the RPS or other automatic

actuation and does not involve a significant

reduction in a margin of safety.

Relocation of the RPS Logic Response Time

value from the TS to the Bases section

involves the use of an alternate regulatory

process for controlling the instrument

response time limit. Any change in the RPS

logic response time value would be evaluated

pursuant to the requirements of 10 CFR

50.59.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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.
17238 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

Northern States Power Company,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request:

November 25, 1996, as supplemented

December 12, 1996.

Description of amendment request:

The proposed amendment would make

changes to Section 2.1.A for the Safety

Limit Minimum Critical Power Ratio

(SLMCPR) and to Section 3.11.C for the

Operating Limit Minimum Critical

Power Ratio (OLMCPR). The proposed

change to Section 2.1.A revises the

SLMCPR value from 1.07 to 1.08 for two

recirculation pump operation and from

1.08 to 1.09 for single loop operation.

The proposed change to Section 3.11.C

deletes the sentence that specifies the

OLMCPR limit penalty for single

recirculation loop operation and adds a

statement that references the Core

Operating Limits Report (COLR) as the

source for this information.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The basis of the MCPR [minimum critical

power ratio] Safety Limit calculation is to

ensure that greater than 99.9% of all fuel rods

in the core avoid transition boiling if the

limit is not violated. The new SLMCPRs

preserve the existing margin to transition

boiling and fuel damage in the event of a

postulated accident. The probability of fuel

damage is not increased. The derivation of

the revised SLMCPRs for Monticello for

incorporation into the Technical

Specification, and its [their] use to determine

cycle-specific thermal limits, have been

performed using NRC-approved methods as

identified in Technical Specification

6.7.A.7.b. NSP [Northern States Power]

methodology established OLMCPR such that

integrity of the SLMCPR is maintained for the

bounding analyzed transients. Additionally,

GENE [General Electric Nuclear Energy]

interim implementing procedures, which

incorporate cycle-specific parameters, have

been used. Based on the use of these

calculations, the calculation of the revised

SLMCPRs maintains the integrity of the

safety limits and therefore cannot increase

the probability or severity of an accident. The

single loop OLMCPR evaluation was

performed using NSP methodology approved

by the NRC. Relocating the OLMCPR value to

the COLR establishes appropriate control on

a core operating limit which may vary from

cycle to cycle because it is cycle dependent.

Since OLMCPR is developed using

procedures approved in the Technical

Specifications, placing the OLMCPR in the

COLR cannot result in a change not

controlled by the Technical Specifications.

The change does not affect failure modes of

equipment, therefore, this amendment will

not cause a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The MCPR Safety Limit is a Technical

Specification numerical value, designed to

ensure that fuel damage from transition

boiling does not occur as a result of the

limiting postulated accident. It cannot create

the possibility of any new type of accident.

The new SLMCPRs have been calculated

using NRC-approved methods and the

OLMCPR values are more conservative.

Additionally, interim procedures, which

incorporate cycle-specific parameters, have

been used. Therefore, the proposed Technical

Specification change does not create the

possibility of a new or different kind of

accident, from any accident previously

evaluated.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

The MCPR Safety Limit is a Technical

Specification numerical value, designed to

ensure that fuel damage from transition

boiling does not occur as a result of the

limiting postulated accident. Increasing the

SLMCPR and OLMCPR values results in an

increase in the margin of safety to fuel

failure, and does not affect other plant

systems. Therefore, the proposed Technical

Specification 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: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Project Director: John N.

Hannon.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 20, 1996, as supplemented by

letter dated February 20, 1997.

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

allow the Vice President to designate

the Safety Audit and Review Committee

(SARC) Chairperson, to change the work

hours limitation in accordance with

guidance in GL 82–12, ‘‘Nuclear Power

Plant Staff Working Hours;’’ to change

radioactive shipments record retention

requirements to comply with recent 10

CFR Part 20 changes; to revise position

titles to reflect organizational changes;

and other editorial changes. The

February 20, 1997, supplemental letter

differs from the November 20, 1996,

application which was noticed in the

Federal Register on January 2, 1997 (62

FR 131), in that the previous application

did not propose changes to TS 5.3, 5.5,

5.6, 5.7, and 5.11 reflecting recent

organizational 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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The changes requested are administrative

in nature. Paragraph 3.D was placed in the

License by Amendment No. 155 to authorize

Omaha Public Power District (OPPD) to

increase the storage capacity of the FCS spent

fuel pool. Amendment No. 155 stated that the

TS as issued would be effective when the last

new rack was installed. Since the last new

rack was installed on August 8, 1994,

Paragraph 3.D is no longer necessary and

should be deleted from the License.

Table of Contents, Section 6.0, ‘‘Interim

Special Technical Specifications,’’

Subsections 6.1 through 6.4 are proposed for

deletion because all of the Specifications

referred to have been deleted by previous

Amendments.

The revision proposed for TS 2.15 (Item 2C

of Table 2–3 & Item 1C of Table 2–4) will

insert the correct terminology (Pressurizer

Low/Low Pressure) into the Functional Unit

description.

The revision proposed for TS 5.2 will

delete the specific working hours as stated

and relocate these requirements to the

Updated Safety Analysis Report (USAR).

Overtime will remain controlled by plant

administrative procedures with the USAR

generally following the guidance of the

NRC’s Policy Statement on working hours

contained in Generic Letter 82–12, ‘‘Nuclear

Power Plant Staff Working Hours.’’

Specifying personnel working hours in TS

does not meet any of the four criteria

contained in 10 CFR 50.36 for inclusion in

the TS. Revisions to plant procedures

containing these requirements are required to

be evaluated in accordance with 10 CFR

50.59. The proposed relocation is similar to

recent Amendments issued to the DavisBesse Nuclear Power Station and the San

Onofre Nuclear Generating Station.

The revision proposed for TS 5.5.2.2 will

replace the specific title of the Chairperson

of the Safety Audit and Review Committee
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17239

and replace it with ‘‘Member as appointed by

the Vice President.’’ This will allow the

flexibility to change chairmanship of the

committee amongst the members.

The revisions proposed to TS 5.3, 5.5, 5.6,

5.7, and 5.11 revise position titles and

reporting responsibilities to reflect

organizational changes. Qualifications for

individuals in these positions meet or exceed

the previous requirements.

The revision to TS 5.10 concerning

retention of records of radioactive shipments

will update the TS to current 10 CFR 20

requirements. Plant procedures already

comply with current 10 CFR 20 record

retention requirements. The addition of the

Section 5.0 title corrects a minor format

discrepancy.

These proposed revisions are

administrative in nature. The proposed

revisions have no effect on any initial

assumptions or operating restrictions

assumed in any accident, nor do these

changes have any effect on equipment

required to mitigate the consequences of an

accident. Therefore the proposed revisions

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.

The proposed revisions correct minor

errors, remove outdated information, are

consistent with changes in organizational

structure, 10 CFR Part 20, or the criteria

contained in 10 CFR 50.36. These changes

will not result in any physical alterations to

the plant configuration, changes to setpoint

values, or changes to the application of

setpoints or limits. No new operating modes

are proposed as a result of these changes.

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.

The revisions listed above correct minor

errors, remove outdated information, or are

consistent with changes in organizational

structure, 10 CFR Part 20, or the criteria

contained in 10 CFR 50.36. These changes

will not result in any physical alterations to

the plant configuration, changes to setpoint

values, or changes to the application of

setpoint or limits. 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Project Director: William H.

Bateman.

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 request: February

26, 1997.

Description of amendment request:

The proposed amendment would revise

the combined Technical Specifications

(TS) for the Diablo Canyon Power Plant,

Unit Nos. 1 and 2 to revise TS 3/4.4.5

and 3.4.6.2, including associated Bases

3/4.4.5 and 3/4.4.6.2, to allow the

implementation of steam generator (SG)

tube voltage based repair criteria for

outside diameter stress corrosion

cracking (ODSCC) indications at tube-totube support plate (TSP) intersections.

The allowed primary-to-secondary

operational leakage from any one SG

would be reduced from 500 gpd to 150

gpd.

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.

Structural Integrity Considerations

The structural criteria ensure that all

indications subjected to voltage-based repair

limits will be able to withstand pressure

loading consistent with the criteria of NRC

Regulatory Guide (RG) 1.121.

Tube burst criteria are inherently satisfied

during normal operating conditions because

of the proximity of the tube support plate

(TSP). It is conservatively assumed that the

entire crevice region is uncovered during the

secondary side blowdown of a main steam

line break (MSLB). Therefore, during a

postulated MSLB accident, tube burst

capability must exceed the RG 1.121 criterion

requiring a margin of 1.43 times the steam

line break pressure differential on tube burst.

Based on the latest industry database, the

RG 1.121 criterion is satisfied by bobbin coil

indications of outside diameter stress

corrosion cracking (ODSCC) with signal

amplitudes less than 8.7 volts. The latest

NRC-approved database will be used for

repair and analysis applications.

Industry testing of model boiler and

operating plant tube specimens for free-span

tubing (no tube support plate (TSP) restraint)

at room temperature conditions show typical

burst pressures in excess of 5,000 psi for

ODSCC indications with voltage

measurements at or below 8.7 volts. This

tube burst capability exceeds the RG 1.121

criterion.

The lower voltage repair limit is

conservatively defined to be 2.0 volts in

accordance with NRC Generic Letter (GL) 95–

05, ‘‘Voltage-Based Repair Criteria for

Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress

Corrosion Cracking,’’ August 3, 1995. This

2.0 volt repair limit is very conservative

because it contains a large safety margin,

based on a structural limit of 8.7 volts. A

maximum allowable upper repair limit (URL)

is also established using the guidance of GL

95–05. The URL is calculated before each

inspection by subtracting the NDE

uncertainty and growth rate allowances from

the current structural limit. The URL for near

term inspections at DCPP Units 1 and 2 is

expected to be about 5.0 volts. Bobbin

indications greater than 2.0 volts and less

than or equal to 5.0 volts that are confirmed

by RPC will be repaired. Bobbin indications

greater than 5.0 volts will be repaired.

Following each inspection, burst

probability analyses are performed for the

end of cycle (EOC) distribution. In

accordance with GL 95–05, the projected

MSLB burst probability must be less than the

threshold value of 1 x 10×2. Based on the

relatively small number and voltages of

ODSCC indications identified to date at

DCPP Units 1 and 2, it is expected that the

near term EOC conditional burst probability

for a faulted SG will be much less than this

threshold value, providing further assurance

of acceptable structural integrity.

Leakage Considerations

PG&E will implement reduced operational

leakage limits as recommended in GL 95–05.

PG&E will revise the TS to implement a

maximum leakage rate of 150 gpd for any one

SG to help preclude the potential for

excessive leakage during power operation in

Modes 1 and 2. The TS has also been

changed to specify that the 150 gpd leak limit

is not necessarily a limiting condition for

operation in Modes 3 and 4. The 150 gpd

leak rate per steam generator has been

established for normal operation. This

leakage rate provides added assurance

against tube rupture at normal and faulted

conditions. In Modes 3 and 4, there is less

differential pressure across the tube and the

potential source term from a tube failure is

much less than in Modes 1 and 2. The

operational leak rate monitoring program is

a defense-in-depth measure that provides a

means for identifying leaks during power

operation to allow for repair before such

leaks can result in tube failure. The leakage

criteria ensure that for indications subjected

to voltage-based repair criteria, induced

leakage under worst-case MSLB conditions

will not result in offsite and control room

dose releases that exceed the applicable

guideline values of 10 CFR 100 and GDC 19.

Relative to the expected leakage during

accident condition loadings, a postulated

MSLB outside of containment, but upstream

of the main steam isolation valve (MSIV),

represents the most limiting radiological

condition for implementation of voltagebased repair criteria. The steam generator

tubes are subjected to an increase in

differential pressure following a MSLB,

resulting in a postulated increase in leakage
17240 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

and associated offsite doses. Leakage

following a MSLB bypasses containment.

PG&E will calculate the primary-tosecondary leakage for degradation subjected

to the voltage repair criteria under worst-case

postulated MSLB conditions. The leak rate

will be compared to the maximum allowable

leak rate limit of 12.8 gpm to ensure that a

postulated MSLB occurring at EOC would

not result in radiological consequences that

are in excess of the applicable offsite and

control room dose guidelines of 10 CFR 100

and GDC 19. Based on the relatively small

number of ODSCC indications identified to

date at DCPP Units 1 and 2, it is expected

that the near term EOC predicted leak rates

for a faulted SG will be much less than the

maximum allowable leak rate limit.

Therefore, based on the structural integrity

and leakage considerations discussed above,

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.

Implementation of the proposed voltagebased repair criteria for ODSCC at TSP

intersections does not introduce any

significant change to the plant design basis.

Use of the criteria does not create a

mechanism which could result in an accident

in the free span because the repair criteria do

not apply to tubes containing ODSCC located

outside the thickness of the TSPs. Based on

the burst probability acceptance limit of

1×10¥2, it is expected that for all plant

conditions, neither a single nor multiple tube

rupture event would likely occur in a steam

generator where voltage-based repair criteria

have been applied.

Steam generator tube integrity is

continually maintained through inservice

inspection and primary-to-secondary leakage

monitoring. Any tubes with ODSCC

degradation in excess of the URL are

repaired.

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.

The use of the bobbin probe to disposition

ODSCC degraded tubes within TSP

intersections by voltage-based repair criteria

is demonstrated to maintain SG tube integrity

in accordance with the requirements of RG

1.121. RG 1.121 describes a method

acceptable to the NRC Staff for meeting GDCs

14, 15, 31, and 32 by reducing the probability

or the consequences of SG tube rupture. This

is accomplished by determining the limiting

conditions of degradation of SG tubing, as

established by inservice inspection, for

which tubes with unacceptable degradation

are removed from service. Upon

implementation of the voltage-based repair

criteria, even under the worst case

conditions, the occurrence of ODSCC at TSP

intersections is not expected to lead to a SG

tube rupture during normal or faulted plant

conditions, nor is it expected to lead to

unacceptable primary-to-secondary leakage.

In addressing the combined effects of a loss

of coolant accident (LOCA) and safe

shutdown earthquake (SSE) on the SGs, as

required by GDC 2, it has been determined

that tube collapse may occur based on

analysis for a large break LOCA plus SSE.

The analysis identifies a maximum of 7.5

percent of tubes per SG located adjacent to

wedge regions that are subject to potential

collapse during combined LOCA and SSE.

Tubes located in the wedge region exclusion

zone will be excluded from application of

voltage-based repair criteria. Thus, existing

tube integrity requirements apply to these

tubes and the margin of safety is not reduced.

Implementation practices using voltagebased repair criteria bounds RG 1.83

considerations. Specifically, GL 95–05

requires the following: (1) enhanced eddy

current inspection guidelines are

implemented to provide consistency in

voltage normalization; (2) 100 percent bobbin

coil inspections are performed each cycle for

all hot leg TSP intersections and all cold leg

TSP intersections down to the lowest cold leg

TSP with known ODSCC indications; and (3)

rotating pancake coil (RPC) inspection of

indications greater than 2 volts are performed

to characterize the principal degradation as

ODSCC. DCPP’s proposed voltage-based

repair criteria implementation practices meet

the above requirements, and in some areas

exceed them (for example, 100 percent

bobbin coil inspections are routinely

performed each cycle on every TSP

intersection).

Implementation of voltage-based repair

criteria at TSP intersections will decrease the

number of tubes which must be repaired.

Since the installation of tube plugs to remove

ODSCC degraded tubes from service reduces

RCS flow margin, voltage-based repair

criteria implementation will help preserve

the margin of RCS flow.

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 request

involves 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.

NRC Project Director: William H.

Bateman.

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 27, 1997.

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant, Unit Nos. 1 and 2

by revising Technical Specifications

(TS) 3/4.8.1.1, ‘‘A.C. Sources—

Operating,’’ to clarify that emergency

diesel generator (EDG) testing is

initiated from standby conditions rather

than ‘‘ambient’’ conditions. The

associated TS Bases will be revised to

discuss the temperature range that

satisfies EDG standby conditions. This

amendment also proposes to revise TS

3/4.3.2, ‘‘Instrumentation—Engineering

Safety Features Actuation System

Instrumentation.’’ This revision clarifies

that when one or both of the first level

load shed relays, or one or both of the

second level undervoltage relays are

inoperable, the associated EDG for that

bus shall be declared inoperable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to the technical

specifications (TS) do not change the

function or operation of any plant equipment

or affect the response of that equipment if it

is called upon to operate.

The proposed change to TS 4.8.1.1.2a.2

and the Bases will clarify the term ‘‘ambient

conditions’’ as used in the emergency diesel

generator (EDG) surveillance requirement.

EDG testing will still be completed on a

frequency commensurate with the current

TS.

The proposed change to TS 3.3.2, Table

3.3–3, will permit time to restore the load

shed first level undervoltage relays (FLURs)

and second level undervoltage relays

(SLURs) to operable status that is consistent

with times allowed for outage of other safetyrelated equipment affecting one train of vital

equipment. This proposed change maintains

a high degree of equipment availability

without requiring unnecessary initiation of a

plant shutdown for partial equipment

outages.

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
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17241

accident from any accident previously

evaluated.

The proposed change to TS 4.8.1.1.2a.2

and the Bases will clarify the term ‘‘ambient

conditions’’ as used in the EDG surveillance

requirement. EDG testing will still be

completed on a frequency commensurate

with the current TS, and will be more

representative of the conditions under which

the EDGs would be required to start in an

accident condition.

The proposed change to TS 3.3.2, Table

3.3–3, will provide time to restore the load

shed FLURs and SLURs to operable status

that is consistent with times allowed for

outage of other safety-related equipment

affecting one train of vital equipment. The

load shed FLUR and SLUR sets for one 4 kV

bus only affect one train of vital equipment.

If an accident occurred while the relays were

inoperable, the redundant trains (two

remaining EDGs and vital buses) would

complete the safety function. The proposed

allowed outage time (AOT) for the load shed

FLURs and SLURs is bounded by the time

allowed for an EDG supporting the vital 4 kV

bus and is consistent with AOTs for other

safety-related components.

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.

The proposed change to TS 4.8.1.1.2a.2

and its Bases, clarifies the term ‘‘ambient

conditions’’ as used in the EDG surveillance

requirement. EDG testing will still be

completed on a frequency commensurate

with the current TS. Use of temperatures in

the standby range result in no significant

variation in EDG start times as indicated by

the diesel vendor and by PG&E test results.

Standby conditions are representative of

actual starting conditions that would be in

effect if the EDGs started in an accident.

The proposed change to TS 3.3.2, Table

3.3–3, will provide time to restore the load

shed FLURs and SLURs to operable status

that is consistent with times allowed for

outage of other safety-related equipment

affecting one train of vital equipment. If an

accident occurred while the relays were

inoperable, the redundant trains (two

remaining EDGs and vital buses) would

complete the safety function. The proposed

change eliminates an unneccessary plant

shutdown and associated risk due to

shutdown transient. It prevents a transient

that could require the EDGs at a time when

less than all three EDGs would be operable.

Therefore, neither of the proposed changes

involves 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.

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

28, 1997.

Description of amendment request:

The proposed amendment by Portland

General Electric (PGE or the licensee)

clarifies the administrative controls that

are used for the revision and

maintenance of the Certified Fuel

Handler Training Program. The change

allows the licensee to make changes to

the certified fuel handlers program

without prior NRC staff approval. The

text of the proposed change is taken

from the improved standard technical

specifications, NUREG–1431, ‘‘Standard

Technical Specifications, Westinghouse

Plants.’’

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, which is presented

below:

In accordance with the requirements of 10

CFR 50.92, ‘‘Issuance of amendment,’’ this

license amendment request is judged to

involve no significant hazards consideration

based upon the following:

1. The proposed license amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is a clarification of

the method of control that will be used for

the Certified Fuel Handler Training Program,

and as such, is administrative in nature and

has no impact on the probability or

consequences of accidents previously

evaluated. The physical structures, systems,

and components of the facility and the

operating procedures for their use are

unaffected by this proposed clarification. The

proposed administrative controls provide

adequate confidence that personnel that

perform the certified fuel handler functions

will have been adequately trained for the

changing conditions of the facility. Since the

training program will prepare the operations

personnel for fuel handling operations,

including responses to abnormal events/

accidents, there will be no increase in the

probability of occurrence or in the

consequences of an accident previously

evaluated.

2. The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

This change ensures the qualifications of

the operations personnel are commensurate

with the tasks to be performed and the

conditions to which they may be required to

respond. This change does not affect plant

equipment or the procedures for operating

plant equipment and, therefore, does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed license amendment does

not involve a significant reduction in a

margin of safety.

This change ensures the qualification of

the operations personnel are commensurate

with the tasks to be performed and the

conditions to which they may be required to

respond. The assumptions for a fuel handling

accident in the Fuel Building are not affected

by the proposed change. The proposed

amendment does not, therefore, involve a

reduction in a margin of safety.

The NRC staff has reviewed the

analysis of the licensee and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Branford Price Millar Library,

Portland State University, 934 S.W.

Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for the Licensees: Leonard A.

Girard, Esq., Portland General Electric

Company, 121 S.W. Salmon Street,

Portland, Oregon 97204.

NRR Project Director: Seymour H.

Weiss.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request:

September 11, 1996.

Description of amendment request:

The proposed amendment would permit

operation with increased safety relief

valve (SRV) and safety valve (SV)

setpoint tolerance and permit operation

up to 100% of rated power with a single

inoperable SRV.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes will permit

operation with increased SRV and SV

setpoint tolerance and permit operation up to

100% of rated power with a single inoperable

SRV.
17242 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

The valves are not related to the control

rod system. The valves are not involved in

the initiation of a Control Rod Drop

Accident. The valves are part of the Reactor

Vessel (RV) pressure boundary and their

failure could initiate a LOCA [loss-of-coolant

accident]. However, the proposed changes do

not constitute a change in the design of the

valves from a pressure boundary perspective.

The proposed changes do not affect the

probability of a LOCA initiated by valve

failure. The valves are not a component,

system, or structure involved in refueling

operations. The valves and their as-found

setpoint tolerance are not involved in the

initiation of a Refueling Accident.

The design basis Main Steam Line Break is

a complete severance of one main steam line

outside the secondary containment. The

SRVs and SVs are located inside primary

containment and cannot cause a main steam

line rupture outside secondary containment.

The valves are not involved in the initiation

of a design basis Main Steam Line Break. The

probability or consequences of these

accidents are not affected.

Attachment C [see application dated

September 11, 1996] includes an analysis to

demonstrate that margin exists to SV

challenges during an Abnormal Operational

Transient (AOT). For this purpose a

Generator Load Rejection without Bypass

(GLRWOBP) was identified as the limiting

AOT. The results confirm that SV challenges

would not occur with an inoperable SRV at

rated power.

The current Technical Specification limit

of 95% rated power or less with an

inoperable SRV is therefore not required to

prevent SV challenges during an AOT.

As discussed in Attachment C [see

application], the impact of the proposed asfound SRV setpoint tolerance increase on

SRV piping/supports and discharge loads to

the Torus was evaluated. A mechanical loads

analysis confirmed the integrity of these

components, systems, and structures during

SRV discharge with the proposed changes.

Attachment C [see application] provides an

evaluation of the impact of the proposed

changes on the consequences of the Loss of

Coolant Accident and the Main Steam Line

Break. The limiting LOCA event is a break in

the recirculation loop, with a break area of

0.6 ft2, at the pump discharge location, with

a loss of one train of DC power as the single

failure. For breaks in the recirculation line

larger than 0.4 ft2, the SRVs would not be

challenged. Therefore, in assessing the

impact of the proposed changes on

10CFR50.46 acceptance criteria, only

recirculation line breaks less than 0.4 ft2

were reevaluated. Results show that the 0.6

ft2 recirculation line break remains the

limiting LOCA event and it is not affected.

The consequences of the limiting design

basis LOCA are not increased by the

proposed changes. The design basis accident

for containment performance is a doubleended break in the recirculation pump

suction. For this size break, the SRVs are not

challenged. Therefore, the proposed changes

do not have any effect on the design basis

accident for containment performance. The

design basis accident for radioactive material

releases and radiological effects is a complete

severance of one main steam line outside the

secondary containment. For steam line

breaks outside the containment, MSIVs [main

steam isolation valves] close and terminate

radiological releases outside the

containment, SRVs are not challenged until

after MSIV closure and isolation. Therefore,

the proposed changes do not increase the

radiological consequences of the design basis

Main Steam Line Break.

The SRVs and SVs are designed to mitigate

the consequences of malfunctions of

equipment which result in a Nuclear System

pressure increase. These abnormal

operational transients are defined and

analyzed in Section 14.5.1 of the VY

[Vermont Yankee] FSAR [final safety analysis

report]. The impact of the proposed changes

on these abnormal operational transients was

evaluated. Results are documented in

Attachment C [see application] and show that

applicable acceptance criteria are met

provided operating MCPR [minimum critical

power ratio] limits as specified in the COLR

[core operating limit report] are adjusted to

reflect the effects of the proposed changes. A

hot channel analysis of the limiting delta

CPR overpressure transient confirmed that a

0.02 increase in the operating MCPR limits

bounds the combined effects of

implementing the proposed changes in the

current cycle. The operating MCPR limits in

COLR have already been increased for the

current cycle. Appropriate operating MCPR

limits for future cycles will be determined

from cycle-specific safety analyses performed

with the approved changes.

Current practice regarding SRV setpoints is

to assure plus or minus 1% tolerance is met

as required by the ASME [American Society

of Mechanical Engineers] Boiler & Pressure

Vessel Code referenced in Technical

Specification Surveillance Requirement

4.6.E.2. As-left setpoints always meet the

plus or minus 1% tolerance. The safety

analysis in Attachment C [see application]

demonstrates that as-found setpoints within

plus or minus 3% are acceptable. However,

valves re-installed after testing will continue,

as previously, to meet plus or minus 1%

tolerance as required by the ASME Boiler &

Pressure Vessel Code. Thus, the probability

of SRV actuation (and the associated risk of

failure to reseat properly) is not increased by

the proposed change.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from an accident previously

evaluated.

The proposed changes will permit

operation with increased Safety Relief Valve

(SRV) and Safety Valve (SV) setpoint

tolerance and permit operation up to 100%

of rated power with a single inoperable SRV.

The proposed changes:

(1) do not constitute a change in the design

of the valves;

(2) will not cause the valve or associated

systems and structures to be operated beyond

their original design envelopes; and,

(3) do not involve new plant equipment.

Therefore, this amendment does not create

the possibility of a new or different kind of

accident.

3. The proposed amendment will not

involve a significant reduction in a margin of

safety.

Technical Specification Basis 3.6 and 4.6D

identifies the minimum critical power ratio

(MCPR) safety limit. Operational restraints on

MCPR are placed in the COLR to assure no

violation of the MCPR safety limit during

AOTs. The impact of the proposed changes

on MCPR limits was determined by

performing a hot channel analysis for the

overpressure transient which yields the

largest transient drop in CPR [critical power

ratio] (delta CPR). Results are documented in

Attachment C [see application], and show

that a 0.02 increase in the operating MCPR

limits bounds the combined effects of the

proposed changes and assures the MCPR

safety limit is not violated during AOTs. The

margin of safety defined by the MCPR safety

limit is not reduced.

Technical Specification Basis 3.6 and 4.6D

also identifies the ASME Boiler and Pressure

Vessel Code Section III-A limit which

permits pressure transients up to 10% over

design pressure (110% x 1250 = 1375 psig).

This margin of safety is not reduced by the

proposed changes. Attachment C [see

application] documents new overpressure

transient analysis with results that

demonstrate the ASME overpressure limit of

110% of design is met. This license

amendment request does not propose to

reduce the margin of safety defined by the

ASME Boiler & Pressure Vessel Code limit.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Ropes and Gray, One International

Place, Boston, MA 02110–2624.

NRC Project Director: Patrick D.

Milano, Acting.

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: February

3, 1997 as supplemented March 18,

1997.

Description of amendment request:

The proposed change to Technical

Specification 4.15.B.1 is administrative

in nature in that it revises the Technical

Specifications (TS) to be consistent with

the NRC-approved inservice inspection

program. In addition, three TS pages

which were previously approved by

NRC, and which were inadvertently

omitted in an earlier amendment

(amendments 40 and 39 for units 1 and

2, respectively), are being reissued.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17243

issue of no significant hazards

consideration, which is presented

below:

1. Operation of Surry Units 1 and 2 in

accordance with the proposed Technical

Specifications change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change is administrative in

nature, and station operations are not being

affected. The ASME Section XI Code

requirements are thoroughly established,

reviewed and approved by ASME, the

industry and ultimately endorsed by the NRC

for inclusion into 10 CFR 50.55a. Updates to

the Code reflect advances in technology and

consider information obtained from plant

operating experience to provide enhanced

inspection and examination techniques for

pipe welds. Therefore, performing weld

examinations for the pipe in our augmented

inspection program to the requirements of

the 1989 edition of the ASME Section XI

Code provides a regulatory acceptable and

adequate level of assurance that the integrity

of the pipe will be maintained. By not

referencing a specific Code edition in the

Technical Specifications, our examinations

for pipe in the augmented inspection

program will consistently be performed to

the Code of record, consistent with the

requirements [of] 10 CFR 50.55a.

Consequently, the probability or

consequences of an accident previously

evaluated are not increased.

2. The proposed Technical Specifications

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

As noted above, the proposed change is

administrative in nature, and no new

accident precursors are being introduced.

Since the augmented inspection program will

continue to be performed to NRC approved

ASME Section XI Code requirements,

adequate assurance is provided to ensure the

integrity of the pipe. Consequently, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed Technical Specifications

change does not involve a significant

reduction in a margin of safety.

Performing weld examinations to the Code

of record is prudent, consistent with

accepted industry and regulatory

requirements, and provides adequate

assurance that piping integrity will be

maintained. The use of a general ASME

Section XI Code reference in Technical

Specification 4.15.B.1 is consistent with the

existing wording in Technical Specifications

4.15.A and C, and ensures that weld

examinations are being consistently

performed to the currently approved edition

of the ASME Section XI Code. This is an

administrative change and as such does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied. The

staff notes that the reissuance of three

TS pages is a purely administrative

matter which involves no significant

hazards consideration and which has

been considered previously. 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.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Dates of amendment requests: June 4,

1996, as supplemented August 5,

September 26, October 21, November

13, November 20, and December 2,

1996, and January 16, March 5, and

March 20, 1997 (TSCR 188 and 189).

Description of amendment requests:

The proposed amendments would

revise License Nos. DPR–24 and DPR–

27 to add commitments for control room

habitability and revise Technical

Specification (TS) Sections 15.1,

‘‘Definitions,’’ 15.2.1, ‘‘Safety Limit,

Reactor Core,’’ 15.2.3, ‘‘Limiting Safety

System Settings and Protective

Instrumentation,’’ Section 15.3.1,

‘‘Reactor Coolant System,’’ 15.3.4,

‘‘Steam and Power Conversion System,’’

15.3.5, ‘‘Instrumentation System, 15.4.1,

‘‘Operational Safety Review,’’ 15.5.3,

‘‘Design Features—Reactor,’’ and 15.6.9,

‘‘Plant Reporting Requirements,’’ and

modify the bases for Section 15.2.2,

‘‘Safety Limit, Reactor Coolant System

Pressure,’’ and Section 15.3.1.C,

‘‘Maximum Coolant Activity,’’ to

incorporate changes associated with the

operation of Point Beach Nuclear Plant

(PBNP), Unit 2, with replacement steam

generators. The new analyses performed

for replacing Unit 2 steam generators

resulted in changes to the reactor core

safety limits and protective

instrumentation setpoints for Unit 1 as

well as Unit 2. Calculations are based on

operation at either 2000 psia or 2250

psia and an average temperature limit of

greater than or equal to 557 degrees

Fahrenheit and less than or equal to

573.9 degrees Fahrenheit. New dose

calculations were performed based on

new setpoints for low-low steam

generator water level, new values of

primary and secondary steam generator

volumes, and revised accident analyses

for steam generator tube rupture, main

steam line break, locked rotor, and

control rod ejection. Additional license

conditions are proposed to document

the commitments made to improve

habitability of the control room so that

dose limits do not exceed 10 CFR Part

50, Appendix A, General Design

Criterion 19, without relying on the use

of potassium iodide pills and/or selfcontained breathing apparatus. The

original applications were previously

noticed in the Federal Register on July

3, 1996 (61 FR 34903 and 34904).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS changes reflect the

replacement of steam generators at PBNP,

including new analyses and setpoints, and a

different standard and acceptance criteria for

Dose Equivalent I–131. The proposed

setpoints maintain the margin to safe

operation of Unit 2 with the replacement

steam generators. In order to maintain one set

of safety analyses for both units, the analyses

for operation of Unit 2 with the replacement

steam generators were performed to

encompass the operation of Unit 1.

Therefore, the proposed changes apply to the

operation of both units and maintain the

margin of safety for each. The staff

independently performed an evaluation of

the dose consequences for steam generator

tube rupture, main steam line break, locked

rotor accident, and a rod ejection accident.

The staff determined there are no significant

increases in dose for the low population zone

or the exclusion area boundary. The licensee

had not previously analyzed these accidents

for control room habitability. As a result of

the proposed changes, limiting control room

doses will require compensatory measures,

use of potassium iodide and self-contained

breathing apparatus, which have been

previously approved, until such time that the

control room ventilation design is improved.

The commitments to improve control design/

operation are included as license conditions.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Installation of new steam generators, with

a small increase in primary side volume and

new setpoints for instrumentation, does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed setpoints
17244 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

maintain the margin to safe operation of Unit

2 with the replacement steam generators. In

order to maintain one set of safety analyses

for both units, the analyses for operation of

Unit 2 with the replacement steam generators

were performed to encompass the operation

of Unit 1. Therefore, the proposed changes

apply to the operation of both units and

maintain the margin of safety for each. These

changes do not affect any of the parameters

or conditions that contribute to initiation of

any accidents. Therefore, the proposed

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

(3) The proposed changes do not involve

a significant reduction in a margin of safety.

The proposed setpoints maintain the

margin to safe operation of Unit 2 with the

replacement steam generators. In order to

maintain one set of safety analyses for both

units, the analyses for operation of Unit 2

with replacement steam generators were

performed to encompass the operation of

Unit 1. Therefore, the proposed changes

apply to the operation of both units and

maintain the margin of safety for each.

Compensatory measures will ensure control

room doses remain within the dose

guidelines in 10 CFR Part 50, Appendix A,

General Design Criterion 19, until such time

as the control ventilation system design/

operation is revised. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

Based on this review, it appears that the

three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards consideration.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John N.

Hannon.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Power Plant, Unit Nos. 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request:

September 30, 1996, as supplemented

November 26, and December 12, 1996,

February 13, and March 5, 1997 (TSCR

192).

Description of amendment request:

The proposed amendments would

revise License Nos. DPR–24 and DPR–

27 to add commitments for control room

habitability and revise Technical

Specification (TS) Sections 15.3.3,

‘‘Emergency Core Cooling System,

Auxiliary Cooling Systems, Air

Recirculation Fan Coolers, and

Containment Spray,’’ TS 15.3.7,

‘‘Auxiliary Electrical Systems,’’ 15.5.2,

‘‘Design Features-Containment,’’ and

associated TS Bases to reflect proposed

changes to the limiting conditions for

operation, action statements, allowable

outage times, and design specifications

for the Point Beach Nuclear Plant

(PBNP) TS associated with the

containment accident fan coolers,

service water equipment (pumps and

piping), component cooling water

pumps, and normal and emergency

power supplies. Specifically, these

proposed changes increase the number

of service water pumps and component

cooling water pumps required to be

operable, change the description of the

service water system to define three

separate loops, modify the limiting

conditions for operation of the

containment cooling and iodine removal

systems and the component cooling

water and service water systems, modify

the auxiliary electrical system

requirements, modify the associated TS

Bases, and change the design value for

each containment ventilation/air coolers

from 55,600 Btu/sec to 41,700 Btu/sec.

The original application was previously

noticed in the Federal Register on

November 19, 1996 (61 FR 58905).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes involve components

currently installed in the facilities and reflect

current capabilities of this equipment.

Increasing the number of service water and

component cooling water pumps required to

be operable, changing the service water

header definitions and modifying the

limiting conditions for operation for service

water and component cooling water, and

modifying the requirements for the 4160/480-

volt safeguards buses does not increase the

probabilities of any accidents currently

evaluated in the final safety analysis report

(FSAR). The probabilities of accidents

previously evaluated in the FSAR are based

on the probability of initiating events for

these accidents. Initiating events for

accidents previously evaluated for Point

Beach include: Control rod withdrawal and

drop, CVCS [chemical volume and control

system] malfunction (boron dilution), startup

of an inactive reactor coolant loop, reduction

in feedwater enthalpy, excessive load

increase, losses of reactor coolant flow, loss

of external electrical load, loss of normal

feedwater, loss of all AC power to the

auxiliaries, turbine overspeed, fuel handling

accidents, accidental releases of waste liquid

or gas, steam generator tube rupture, steam

pipe rupture, control rod ejection, and

primary coolant system ruptures. The change

to the heat removal capability of the

containment ventilation/air coolers from

55,600 Btu/sec to 41,700 Btu/sec was

evaluated to ensure that containment design

is not challenged. Therefore, the proposed

changes do not affect the probability of

occurrence or the consequences of any

accident previously evaluated in the FSAR.

During review of the proposed changes, the

staff determined that other changes made to

the operation of the containment spray

system and the control room ventilation

design and operation could affect the doses

associated with a loss-of-coolant accident.

The staff has determined that there is no

significant increase in offsite doses. As a

result of the proposed changes and current

plant design, limiting control room doses

will require compensatory measures, use of

potassium iodide and self-contained

breathing apparatus, which have been

previously approved, until such time that the

control room ventilation design/operation is

improved. The commitments to improve

control design/operation are included as

license conditions.

(2) The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not introduce

any new accidents from any previously

evaluated. Failures for the systems affected

by the proposed changes, service water

system, component cooling water system,

containment ventilation/air cooling units,

and the 4160/480-volt safeguards buses are

factored into the accident analyses included

in the FSAR. No new or different kinds of

accidents are created since no new or

different accident initiators or sequences are

involved. Therefore, these proposed changes

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated in the Point Beach

FSAR.

(3) The proposed changes do not involve

a significant reduction in a margin of safety.

The proposed changes provide the

appropriate limiting conditions for operation,

action statements, allowable outage times,

and design specifications for service water,

component cooling water, containment

cooling, and normal and emergency power

supplies. This ensures that the safety systems

that protect the reactor and containment will

operate as required. The impact of changes to

design and operation of affected systems do

not affect the reactor and containment

design. Therefore, the margins of safety for

Point Beach are not being reduced because

the design and operation of the reactor and

containment are not being changed and the

safety systems and limiting conditions of

operation for these safety systems that

provide their protection that are being

changed will continue to meet the

requirements for accident mitigation for

PBNP. Compensatory measures will ensure

control room doses remain within the dose

guidelines in 10 CFR Part 50, Appendix A,

General Design Criterion 19, until such time

as the control ventilation system design/
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17245

operation is revised. Therefore, the proposed

changes will not create a significant

reduction in a margin of safety.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John N.

Hannon.

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.

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: March 7,

1997.

Description of amendment request:

The proposed amendments would

revise Technical Specification 3/4.7.1.6

and Section 15.6.3 of the Updated Final

Safety Analysis Report to require four

instead of three steam generator

pressure operated relief valves operable.

Date of publication of individual

notice in Federal Register: March 13,

1997 (62 FR 11931).

Expiration date of individual notice:

April 14, 1997.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388 Susquehanna Steam Electric

Station, Unit 2, Luzerne County,

Pennsylvania

Date of amendment request: March

17, 1997.

Brief description of amendment

request: The proposed amendment

would modify the Design Features

Section 5.3.1 of the Technical

Specifications to reflect the Atrium-10

design and would include a Siemens

Power Corporation topical report

reference in Section 6.9.3.2 to reflect

mechanical design criteria for this fuel.

This change would allow this fuel to be

loaded and maintained in the core only

under Condition 5, (refueling).

Date of publication of individual

notice in Federal Register: March 25,

1996 (62 FR 14167).

Expiration date of individual notice:

April 24, 1997.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) The applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

May 2, 1995, as supplemented by letter

dated March 7, 1996.

Brief description of amendments:

These amendments modify the licenses

to authorize incorporation in the

Updated Final Safety Analysis Report

(UFSAR) of certain changes to the

description of the facilities involving a

revised large-break loss of coolant

accident (LOCA) analysis that addresses

a previously unanalyzed release path

through the steam generators to the

atmosphere.

Date of issuance: March 17, 1997.

Effective date: March 17, 1997, to be

implemented within 60 days of

issuance.

Amendment Nos.: Unit 1—111; Unit

2—103; Unit 3—83.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Operating

Licenses and Updated Final Safety

Analysis Report.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62487). The March 7, 1996,

supplemental letter provided additional

clarifying information and did not

change the initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated March 17, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

January 29, 1997, as supplemented

February 6, and February 21, 1997.

Brief description of amendment: The

amendment adds a new Technical
17246 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

Specification 3.0.5 to provide guidance

for returning equipment to service

under administrative controls for the

sole purpose of performing testing to

demonstrate operability.

Date of issuance: March 17, 1997.

Effective date: March 17, 1997.

Amendment No.: 69.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62FR6569).

The February 6, and February 21,

1997 letters provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 17, 1997.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

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:

January 10, 1997, as supplemented

January 31, February 20, and March 3,

1997.

Brief description of amendment: The

amendment revises Technical

Specification 4.8.1.1.2 to clarify

pressure testing requirements for the

isolable and non-isolable portions of the

diesel fuel oil piping.

Date of issuance: March 19, 1997.

Effective date: March 19, 1997.

Amendment No.: 70.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 5, 1997 (62 FR 5490).

The January 31, February 20, and March

3, 1997, letters provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 19, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina.

Date of application for amendments:

November 4, 1996 and supplemented

February 5, 1997.

Brief description of amendments: The

amendments revise Technical

Specification Section 4.7.13.1.c to

eliminate the requirement that the 18-

month Standby Shutdown System

diesel generator inspection be

performed only during shutdown of

both reactors.

Date of issuance: March 13, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1—157—Unit

2—149.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64383) The supplemental letter dated

February 5, 1997, provided additional

information that did not change the

scope of the November 4, 1996,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 13, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

January 13, 1997.

Brief description of amendments: The

amendments revise the Technical

Specifications so that the containment

integrated leak rate Type A testing will

now be performed consistent with the

revised 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leak-Test Program.’’ No

changes to implement Option B for the

Type B and Type C tests were requested

by the licensee at this time.

Date of issuance: March 21, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1—173—Unit

2—155.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6575) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 21, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, North Carolina 28223–0001.

Duke Power Company, Docket Nos. 50–

269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

February 15, 1996, as supplemented by

letter dated February 18, 1997.

Brief description of amendments: The

amendments add operability and

surveillance requirements regarding

operation and testing of the Keowee

Hydro Station to the Oconee Technical

Specifications.

Date of Issuance: March 20, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days. Implementation shall include

revision of the Selected Licensee

Commitment manual to incorporate the

Keowee Hydro units’ commercial power

operating restrictions curves in

accordance with the application for the

amendments.

Amendment Nos.: Unit 1—222; Unit

2—222; Unit 3—219.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13523)

The February 18, 1997, letter provided

clarifying information that did not

change the scope of the February 15,

1996, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 20, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17247

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–321 and 50–

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

September 18, 1992, as supplemented

October 6, 8, 15, 23, and November 13

and 20, 1992, March 5, May 24, June 10,

and December 20, 1993, April 6 and July

28, 1995, and September 11, October 1,

December 13, 19 and 23, 1996.

Brief description of amendments: The

amendments modify the Facility

Operating Licenses, Technical

Specifications, Environmental

Protection Plan, and Antitrust

conditions to add Southern Nuclear

Operating Company, Inc., as operator of

the facilities, with exclusive

responsibility and control over its

physical construction, operation, and

maintenance. The antitrust license

conditions divorce Southern Nuclear

from marketing or brokering power or

energy from the Hatch Plant and holds

Georgia Power Company accountable for

the actions of its agent, Southern

Nuclear, to the extent Southern

Nuclear’s actions contravene the Hatch

antitrust license conditions. An Order

Approving Southern Nuclear Operating

Company, Incorporated, As Exclusive

Operator was included along with the

issuance of the amendments.

Date of issuance: March 17, 1997.

Effective date: To be implemented

within 60 days of the date of issuance.

Amendment Nos.: 203 and 144.

Facility Operating License Nos. DPR–

57 and NPF–5: Amendments revised the

Technical Specifications and Operating

Licenses.

Date of initial notice in Federal

Register: October 14, 1992 (57 FR

47131). The October 6, 8, 15, 23, and

November 13 and 20, 1992, March 5,

May 24, June 10, and December 20,

1993, April 6 and July 28, 1995, and

September 11, October 1, December 13,

19 and 23, 1996, letters, did not change

the scope of the September 18, 1992,

application and the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

March 17, 1997, and an Environmental

Assessment dated October 27, 1992.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–321 and 50–

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

October 7, 1996.

Brief description of amendments: The

amendments revise Surveillance

Requirements (SRs) 3.1.7.7 and 3.4.3.1,

and Limiting Conditions for Operation

3.4.3, 3.5.1, and 3.6.1.6 to increase the

nominal mechanical pressure relief

setpoints for all of the 11 safety/relief

valves (SRVs) to 1150 psig and allow

operation with one SRV and its

associated functions inoperable. The

change will reduce the potential for SRV

pilot leakage and the potential for forced

outages due to an inoperable SRV

during a fuel cycle.

Date of issuance: March 21, 1997.

Effective date: As of the date of

issuance to be implemented for Unit 1

prior to startup from its refueling outage

scheduled for fall 1997; and for Unit 2

prior to startup from its refueling outage

currently scheduled for March 15, 1997.

Amendment Nos.: 204 and 145.

Facility Operating License Nos. DPR–

57 and NPF–5: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 129).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 21, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–321 and 50–

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

October 29, 1996, as supplemented

February 19, 1997.

Brief description of amendments: The

amendments revise the Technical

Specifications associated with the

installation of a digital Power Range

Neutron Monitoring system.

Date of issuance: March 21, 1997.

Effective date: As of the date of

issuance to be implemented for Unit 1

prior to its startup from the fall of 1997

refueling outage; and implemented for

Unit 2 prior to its startup from the

spring of 1997 refueling outage.

Amendment Nos.: 205 and 146.

Facility Operating License Nos. DPR–

57 and NPF–5: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 130).

The February 19, 1997, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 21, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear

Date of application for amendment:

November 27, 1996 (TSCR 232).

Brief description of amendment: The

amendment changes the acceptance

criteria for the individual cell voltage

from 2.0v to 2.09v, the frequency for

battery specific gravities to implement

the recommendations of IEEE 450–1995,

deletes surveillance 4.7.B.4.d, and adds

a clarifing phrase ‘‘while on a float

charge . . .’’ where appropriate.

Date of Issuance: March 24, 1997.

Effective date: March 24, 1997.

Amendment No.: 189.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6576) The Commission’s related

evaluation of this amendment is

contained in a Safety Evaluation dated

March 24, 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.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 20, 1996, as supplemented by

letters dated February 20, 1997, and

March 25, 1997.

Brief description of amendment: The

amendment revises Section 5.2 of the

Fort Calhoun Station technical

specifications to relocate controls for

working hours to the Updated Safety

Analysis Report.

Date of issuance: March 27, 1997.

Effective date: March 27, 1997.

Amendment No.: 181.
17248 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

Facility Operating License No. DPR–

40: Amendment revised the Technical

Specifications and operating license.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 131)

The February 20, 1997, and March 25,

1997, supplemental letters provided

additional clarifying information that

did not change the portion of the initial

no significant hazards consideration

determination that addressed this

proposed change.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 27, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

January 11, 1996, as supplemented by

letters dated February 26, May 22, June

27, July 12, December 23, 1996, and

March 17, 1997

Brief description of amendment: The

amendments revise Section 6.0

(Administrative Controls) of the Hope

Creek TS to: (1) Relocate the

requirements of Section 6.5 (Station

Operations Review Committee, Nuclear

Safety Review and Audit, and Technical

Review and Control) to the Quality

Assurance Program, (2) replace specific

management titles with generic

management functional positions, (3)

change Operating Engineer to Assistant

Operations Manager, (4) require a Senior

Reactor Operator license be held by

either the Operations Manager or one of

the Assistant Operations Managers, and

(5) correct some typographical errors in

Section 6.0.

Date of issuance: March 21, 1997.

Effective date: As of date of issuance

and shall be implemented within 60

days.

Amendment No.: 97.

Facility Operating License No. NPF–

57: This amendment revised the

Technical Specifications and the

license.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5817).

The supplemental letters provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination nor the original notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 21, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

October 25, 1996, as supplemented by

letters dated December 4, 1996, and

January 24, 1997.

Brief description of amendment: This

amendment changes Hope Creek

Technical Specification (TS) 3/4.1.3.5,

‘‘Control Rod Scram Accumulator,’’ in

order to: 1) Permit a separate entry into

a TS action statement for each

inoperable control rod; 2) provide more

specific applicability for required

actions in Operational Condition 1 or 2

with one inoperable control rod scram

accumulator (reactor pressure of ≥ 900

psig would be specified); 3) provide

more specific actions (verify charging

water pressure) for two or more

inoperable control rod scram

accumulators when reactor pressure is ≥

900 psig; 4) provide more specific

actions when reactor pressure is < 900

psig and one or more control rod scram

accumulators are inoperable (verify

insertion of control rods associated with

inoperable accumulators and verify that

charging water header pressure is ≥ 940

psig); 5) provide specific actions in

Operational Condition 5 with one or

more withdrawn control rods

inoperable; and 6) eliminate the

requirements to perform an 18-month

channel functional test of the leak

detectors and the 18-month channel

calibration of the pressure detectors.

Date of issuance: March 26, 1997.

Effective date: As of date of issuance,

to be implemented within 60 days.

Amendment No.: 98.

Facility Operating License No. NPF–

57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64394) The December 4, 1996, and

January 24, 1997, supplements did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

March 26, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

January 11, 1996, as supplemented by

letters dated February 26, May 22, June

27, July 12, December 23, 1996, and

March 17, 1997.

Brief description of amendments: The

amendments revise Section 6.0

(Administrative Controls) of the Salem

TS to: 1) relocate the requirements of

Section 6.5 (Station Operations Review

Committee, Nuclear Safety Review and

Audit, and Technical Review and

Control) to the Quality Assurance

Program, 2) replace specific

management titles with generic

management functional positions, 3)

change Operating Engineer to Assistant

Operations Manager, 4) require a Senior

Reactor Operator license be held by

either the Operations Manager or one of

the Assistant Operations Managers, and

5) correct some typographical errors in

Section 6.0.

Date of issuance: March 21, 1997.

Effective date: Both units, as of date

of issuance, to be implemented within

60 days.

Amendment Nos.: 192 and 175.

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments

revised the Technical Specifications and

the license.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5818) The supplemental letters

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination nor the original notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 21, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 Joseph M.

Farley Nuclear Plant, Unit 1, Houston

County, Alabama

Date of amendment request:

December 26, 1997, as supplemented by

letter dated February 6, March 7, and

March 21, 1997.

Brief Description of amendment: The

amendment changes Technical

Specification 3/4.4.6, ‘‘Steam

Generators’’ and associated Bases to

implement the voltage-based alternate

repair criteria for steam generator tubes

in Farley Unit 1 in accordance with
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17249

Generic Letter 95–05, ‘‘Voltage-Based

Repair Criteria for Westinghouse Steam

Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.’’

Date of issuance: March 24, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 124.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4353)

By letter dated February 6, 1997, the

licensee submitted additional

information to clarify the changes to the

proposed repair criteria, which did not

change the scope of the December 26,

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 24, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request: January

10, 1997, as supplemented by letter

dated February 24, 1997.

Brief Description of amendments: The

amendments revise the Technical

Specifications (TS) to incorporate the

latest revised topical reports governing

the installation of laser welded steam

generator tube sleeves. In addition, the

reference to a one-cycle implementation

of L*, which expired at the last Unit 2

outage was deleted from the Unit 2 TS.

Date of issuance: March 24, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 125 and 119.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4355)

The February 24, 1997, letter provided

clarifying information that did not

change the original application and the

initial proposed no significant hazards

consideration determination published

in the Federal Register on January 29,

1997 (62 FR 4355).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 24, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request:

September 30, 1996.

Brief description of amendments: The

amendments revise Technical

Specifications (TS) 3/4.1.1.1, 3/4.1.1.2,

3/4.1.1.3, 3/4.1.3.5, 3.1.3.6, 3.2.1, 3.2.2

and 3.2.3 and associated Bases to

remove certain cycle-specific parameter

limits from the TS and relocate them to

the Core Operating Limits Report.

Date of issuance: March 25, 1997.

Effective date: As of the date of

issuance to be implemented for Unit 1

prior to entry into Mode 5 following the

next scheduled refueling outage, which

should begin in March 1997; for Unit 2

prior to entry into Mode 5 following the

refueling outage scheduled to begin in

March 1998.

Amendment Nos.: 126 and 120.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications and License

Conditions.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57491) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 25, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

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 18, 1997, as supplemented by

letter dated February 21, 1997.

Brief description of amendment: The

amendment defers implementation of

Surveillance Requirement 3.3.5.6 of

Technical Specifcation 3.3.5,

‘‘Engineered Safety Features Actuation

System (ESFAS) Instrumentation,’’ 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 17, 1997.

Effective date: March 17, 1997.

Amendment No.: 127

Facility Operating License No. NPF–

15: The amendments revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (62 FR 9001 dated

February 27, 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 31, 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

February 21, 1997, letter provided

additional clarifying information and

did not change the original no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

March 17, 1997.

Attorney for licensee: Douglas K.

Porter, 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.

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 application for amendment:

February 14, 1997.

Brief description of amendment: This

amendment revises Technical

Specification (TS) Section 3/4.5.2,

‘‘Emergency Core Cooling Systems,

ECCS Subsystems—Tavg ≥ 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.

This amendment also revises the

applicable TS bases.

Date of issuance: March 24, 1997.

Effective date: Immediately, and shall

be implemented no later than 120 days

after issuance.

Amendment No.: 215.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (62 FR 8783

dated February 26, 1997). The notice
17250 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by March 30, 1997,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of NSHC are contained in a Safety

Evaluation dated March 24, 1997.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

United States Department of Commerce,

National Institute of Standards and

Technology, Docket No. 50–184, NIST

Test Reactor

Date of application for amendment:

January 17, 1997.

Brief description of amendment: This

amendment revises the Technical

Specifications to change the name of the

Reactor Radiation Division to the NIST

Center for Neutron Research and the

Chief, Radiation Division to Director,

NIST Center for Neutron Research.

Date of issuance: March 31, 1997.

Effective date: March 31, 1997.

Amendment No.: 8.

Amended Facility License No. TR–5:

This amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 26, 1997 (62 FR

8801). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 31, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

Location: N/A.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) The application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By May

9, 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
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17251

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–001, 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 20555–001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

December 27, 1996, as supplemented by

letter dated March 18, 1997.

Brief description of amendments: The

amendments modify the licenses to

authorize incorporation of certain

changes to the description of the

facilities involving offsite power sources

into the Updated Final Safety Analysis

Report (UFSAR) for the Palo Verde

Nuclear Generating Station (PVNGS).

Date of issuance: March 26, 1997.

Effective date: March 26, 1997, to be

implemented within 60 days of the date

of issuance.

Amendment Nos.: Unit 1—112; Unit

2—104; Unit 3—84.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the operating

licenses and the Updated Final Safety

Analysis Report.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendments, finding of

emergency circumstances, and final

determination of no significant hazards

consideration are contained in a Safety

Evaluation dated March 26, 1997.

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004.

NRC Project Director: William H.

Bateman.

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 application for amendments:

March 26, 1997, as supplemented on

March 27, 1997.

Brief description of amendments: The

proposed amendments provided (1) An

evaluation of the Unreviewed Safety

Question (USQ) involving the control

room operator dose resulting from error

in the secondary containment volume,

(2) a change in Technical Specification

(TS) 4.7.P.2.b and 4.7.P.3 values for the

allowed methyl iodide penetration for

the standby gas treatment charcoal

adsorbers, and (3) change of TS 5.2.C to

reflect the new calculated free volume

of the secondary containment.

Date of Issuance: March 27, 1997.

Effective date: March 27, 1997.

Amendment Nos.: 175, 171.

Facility Operating License Nos. DPR–

29 and DPR–30: The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendments, finding of

emergency circumstances and final

determination of no significant hazards

consideration are contained in a Safety

Evaluation dated March 27, 1997.
17252 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

NRC Project Director: Robert A. Capra.

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

January 29, 1997, as supplemented

February 11, 12, March 7, 10, 11, 19,

and 20, 1997.

Brief description of amendments: The

amendments authorize Northern States

Power Company to continue operation

of Prairie Island Units 1 and 2 on an

interim basis, through the incorporation

of three license conditions into its

licenses, until a seismically qualified

emergency cooling water source is

provided that will provide the basis to

extend the time for operator postseismic cooling water load management.

This could be done either through a

seismic evaluation of the intake canal,

physical modifications to the intake

canal or plant, or some combination of

the two.

Date of issuance: March 25, 1997.

Effective date: March 25, 1997, with

implementation of License Condition 1

prior to Unit 2 entering Mode 2, with

implementation of the requirements of

License Condition 2 by July 1, 1997, and

December 31, 1998, and with

implementation of License Condition 3

at the next updated safety analysis

report update following completion of

License Condition 2, but no later than

June 1, 1999.

Amendment Nos.: 128 and 120.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the licenses.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (62 FR 5857

dated February 7, 1997). This notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. The

notice also provided for an opportunity

to request a hearing by March 10, 1997,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendments. Because of

the significant revisions to the licensee’s

original application, NRC also

published a public notice of the

proposed amendments, issued a

proposed finding of no significant

hazards consideration, and requested

that any comments on the proposed

finding be provided to the staff by close

of business on March 20, 1997. The

notice was published in the St. Paul

Pioneer Press on March 15, 1997, the

Minneapolis Star Tribune on March 16,

1997, and the Red Wing Republican

Eagle on March 17, 1997. No comments

have been received. The Commission’s

related evaluation of the amendments,

finding of exigent circumstances, and

final determination of NSHC are

contained in a Safety Evaluation dated

March 25, 1997.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

NRC Project Director: John N.

Hannon.

Dated at Rockville, Maryland, this 2nd day

of April, 1997.

For the Nuclear Regulatory Commission.

Jack W. Roe,

Director, Division of Reactor Projects III/IV,

Office of Nuclear Reactor Regulation.

[FR Doc. 97–8916 Filed 4–8–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–219, License No. DPR–16]

Oyster Creek Nuclear Generating

Station; Issuance of Final Director’s

Decision Under 10 CFR 2.206

Notice is hereby given that the

Director, Office of Nuclear Reactor

Regulation, U.S. Nuclear Regulatory

Commission (NRC), has granted in part

and denied in part Petitions, dated

September 19, 1994, and supplemented

by a letter dated December 13, 1994,

submitted by Messrs. Paul Gunter and

William deCamp, Jr. (Petitioners) on

behalf of Oyster Creek Nuclear Watch,

Reactor Watchdog Project, and Nuclear

Information and Resource Service.

Petitioners requested that the NRC take

immediate action with regard to Oyster

Creek Nuclear Generating Station

(OCNGS) operated by GPU Nuclear

Corporation (GPU or licensee). By letter

dated December 13, 1994, Petitioners

supplemented the Petition dated

September 19, 1994.

Specifically, the Petition of September

19, 1994, requested that the NRC (1)

immediately suspend the OCNGS

operating license until the licensee

inspects and repairs or replaces all

safety-class reactor internal component

parts subject to embrittlement and

cracking, (2) immediately suspend the

OCNGS operating license until the

licensee submits an analysis regarding

the synergistic effects of through-wall

cracking of multiple safety-class

components, (3) immediately suspend

the OCNGS operating license until the

licensee has analyzed and mitigated any

areas of noncompliance with regard to

irradiated fuel pool cooling as a singleunit boiling water reactor (BWR), and

(4) issue a generic letter requiring other

licensees of single-unit BWRs to submit

information regarding fuel pool boiling

in order to verify compliance with

regulatory requirements, and to

promptly take appropriate mitigative

action if the unit is not in compliance.

The supplemental Petition, in

addition to providing more information

on the original request, requested that

the NRC (1) suspend the OCNGS

operating license until the Petitioners’

concerns regarding cracking are

addressed, including inspection of all

reactor vessel internal components and

other safety-related systems susceptible

to intergranular stress-corrosion

cracking and completion of any and all

necessary repairs and modification; (2)

explain the discrepancies between the

response of the NRC staff dated October

27, 1994, to the Petition of September

19, 1994, and time-to-boil calculations

for the FitzPatrick plant; (3) require GPU

to produce documents for evaluation of

the time-to-boil calculation for the

OCNGS irradiated fuel pool; (4) identify

redundant components that may be

powered from onsite power supplies to

be used for spent fuel pool cooling as

qualified Class IE systems; (5) hold a

public meeting in Toms River, New

Jersey, to permit presentation of

additional information related to the

Petition; and (6) treat the Petitioners’

letter of December 13, 1994, as a formal

appeal of the denial of their request of

September 19, 1994, to immediately

suspend the OCNGS operating license.

The Director of the Office of Nuclear

Reactor Regulation has granted requests

(3), with the exception of suspending

OCNGS operating license which was

previously denied, and in part (4) of the

Petition of September 19, 1994, and

requests (2), (3), and (4) of the

supplemental Petition of December 13,

1994. The reasons for these decisions

are explained in the ‘‘Final Director’s

Decision Under 10 CFR 2.206: (DD–97–

08), the complete text of which follows

this notice. The decision and the

documents cited in the decision are

available for public inspection and

copying at the Commission’s Public