28604 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

is available for public inspection at the

Commission’s Public Document Room,

the Gelman Building, 2120 L Street,

NW., Washington, DC, and at the local

public document room located at the

White Plains Public Library, 100

Martine Avenue, White Plains, New

York 10601.

Dated at Rockville, MD, this 31st day of

May 1996.

For the Nuclear Regulatory Commission.

George F. Wunder,

Project Manager, Project Directorate I–1,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 96–14044 Filed 6–4–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from May 11,

1996, through May 23, 1996. The last

biweekly notice was published on May

22, 1996 (61 FR 25696).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards onsideration 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 July 5, 1996, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

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
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28605

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: May 1,

1996

Description of amendment request:

The proposed amendment will relocate

the administrative controls related to

the quality assurance review and audit

requirements of Section 6 from the

Pilgrim Station Technical Specifications

to the Boston Edison Quality Assurance

Manual. This change is in accordance

with the guidance contained in NRC

Administrative Letter 95-06,

‘‘Relocation of Technical Specification

Administrative Controls Related to

Quality Assurance.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The change will relocate the administrative

controls related to the quality assurance

review and audit requirements from the

technical specifications to the quality

assurance plan. These changes are

administrative in nature and do not impact

initiators of analyzed events, accident

mitigation capabilities, or transient events.

The quality assurance program is a logical

candidate for such relocation due to the

controls imposed by such regulations as

Appendix B to 10 CFR [Part] 50, the

existence of NRC approved quality assurance

plans and commitments to industry quality

assurance standards, and the established

quality assurance program change control

process in 10 CFR 50.54(a). Therefore, the

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The change will relocate the administrative

controls related to the quality assurance

review and audit requirements from the

technical specifications to the quality

assurance plan. The quality assurance

program is a logical candidate for such

relocation due to the controls imposed by

such regulations as Appendix B to 10 CFR

[Part] 50, the existence of NRC approved

quality assurance plans and commitments to

industry quality assurance standards, and the

established quality assurance program

change control process in 10 CFR 50.54(a).

The proposed changes do not involve a

physical alteration of the plant or changes in

methods governing plant operation. The

changes will not impose or eliminate any

new or different requirements. Therefore the

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The change will relocate the administrative

controls related to the quality assurance

review and audit requirements from the

technical specifications to the quality

assurance plan. These changes are

administrative in nature. The quality

assurance program is a logical candidate for

such relocation due to the controls imposed

by such regulations as Appendix B to 10 CFR

[Part] 50, the existence of NRC approved

quality assurance plans and commitments to

industry quality assurance standards, and the

established quality assurance program

change control process in 10 CFR 50.54(a).

The proposed change will not reduce a

margin of safety because it has no impact on

any safety analysis assumptions. Therefore,

the operation of PNPS [Pilgrim Nuclear

Power Station] in accordance with the

proposed license amendment will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Jocelyn A.

Mitchell, Acting
28606 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: May 1,

1996

Description of amendment request:

The proposed amendment will reflect

the implementation of 10 CFR Part 50,

Appendix J, Option B at the Pilgrim

Nuclear Power Station.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The proposed changes do not

involve any physical or operational changes

to structures, systems or components. The

proposed changes provide a mechanism

within the TS [Technical Specifications] for

implementing a performance-based leakage

rate test program which was promulgated by

the revision to 10CFR50 to incorporate

Option B into Appendix J. The TS Limiting

Conditions for Operation (LCO) remain

unaffected by these changes. Thus, the safety

design basis for the accident mitigation

functions of the primary containment is

maintained. Therefore, these changes will not

increase the probability or consequences of

an accident previously evaluated.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Revising surveillance requirement

acceptance criteria and frequencies does not

physically modify the plant and does not

modify the operation of any existing

equipment. Further, the TS LCOs remain

unaffected by these changes.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety.

The proposed changes do not involve a

significant reduction in the margin of safety,

nor do they affect a safety limit, an LCO, or

the manner in which plant equipment is

operated. The NRC letter dated November 2,

1995, recognizes that changes similar to the

proposed changes are required to implement

Option B of 10CFR50, Appendix J. In

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’ which

forms the basis for the Appendix J revision,

the NRC concludes that adoption of

performance-based test intervals for

Appendix J testing will not significantly

reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Jocelyn A.

Mitchell, Acting

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: May 1,

1996

Description of amendment request:

The proposed amendment would

modify the definition of ‘‘Core

Alteration,’’ and the Limiting Condition

for Operation, Surveillance conditions

and Bases section associated with

Technical Specification (TS) 3.7.C,

‘‘Secondary Containment.’’

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.

Operation of PNPS [Pilgrim Nuclear Power

Station] in accordance with the proposed

license amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because of the following:

Proposed Change ι1: Definition of

‘‘Alteration of the Reactor Core≥

The definition, ‘‘Alteration of the Reactor

Core’’, is being revised so that the term will

apply only to those activities that create the

potential for a reactivity excursion and,

therefore, warrant special precautions or

controls in the TS. The proposed definition

includes normal control rod movement in the

definition, but excludes control rod drive

movement (such as rod removal from the

core) when all four fuel bundles surrounding

a control rod are removed. The proposed

change does not increase the probability or

consequences of an accident because the

proposed definition, by identifying activities

with the potential for causing a reactivity

excursion, ensures that the additional

precautions and controls in the TS are

implemented at all appropriate times. In

addition, the movement of components

excluded by this definition is not assumed in

the initiation of any analyzed event.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Proposed Change ι2: Secondary

Containment

The current specifications are revised to

specify more clearly when secondary

containment is required, what actions to take

if secondary containment is inoperable, and

time frames for completing the actions. These

revisions enhance the existing specification

and serve to make it more definitive by

encompassing the conditions currently

specified by TS and supplementing them to

specify other conditions when secondary

containment is required.

Surveillances 4.7.C.1.a and b were only

necessary during initial and Cycle 1

operations. Removing obsolete information

from the existing specifications, renumbering and re-arranging the wording is

an administrative change.

These changes are administrative in nature

and do not impact initiators of analyzed

events, accident mitigation capabilities, or

transient events. Therefore, the changes do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The operation of PNPS in accordance with

the proposed license amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because of the

following:

Proposed Change ι1: Definition of

‘‘Alteration of the Reactor Core≥

The definition change specifies more

accurately which component movements

constitute a ‘‘Core Alteration’’. This change

does not involve a physical alteration of the

plant (no new or different type of equipment

will be installed) or changes in methods

governing normal plant operation. The

proposed changes will allow movement of

some components (camera, lights, etc.)

during times when ‘‘Core Alterations’’ have

been halted since these components will not

affect core reactivity. Removal of a control

rod involves unlatching and withdrawal/

insertion from over-vessel handling

equipment. These activities necessitate, by

design, the removal of the adjacent four fuel

assemblies. With this configuration (no fuel

in the cell; handling the associated control

rod), the proposed change will allow

movement of a ‘‘reactivity control

component’’ while not imposing

requirements unique to ‘‘Core Alterations’’

(note: other requirements, such as those for

handling loads over irradiated fuel, will

remain applicable). The reactivity effects of

this control rod movement are more than

compensated for by the initial removal of the

fuel assemblies. Therefore, this change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Proposed Change ι2: Secondary

Containment

The proposed change does not eliminate or

relax any existing TS condition. Rather, it

better defines when secondary containment

is required, provides action statements for

inoperability and removes obsolete
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28607

requirements (from first operating cycle).

This change does not involve a physical

change to structures, systems or components,

and the safety design bases for the accident

mitigating function of the secondary

containment is maintained. Therefore, these

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The operation of PNPS in accordance with

the proposed license amendment will not

involve a significant reduction in a margin of

safety because of the following:

Proposed Change ι1: Definition of

‘‘Alteration of the Reactor Core≥

The proposed definition more accurately

identifies those activities with the potential

for causing a reactivity excursion. The more

accurate identification of ‘‘Core Alterations’’

will ensure that when there is a potential for

reactivity excursions, appropriate

precautions are applied. The components

now excluded from the proposed definition

are those that do not have the capability for

adversely impacting core reactivity. The

proposed change has no impact on safety

analysis assumptions. Therefore, the change

will not involve a significant reduction in a

margin of safety.

Proposed Change ι2: Secondary

Containment

The proposed additions of applicability

conditions provide a more precise

understanding of when secondary

containment integrity is required and what

actions to take if it becomes inoperable. The

change does not eliminate any existing

conditions. The deletion of surveillances

applicable only for the first operating cycle

and re-numbering and re-arranging the

remaining surveillance wording is an

administrative change and has no impact on

the operation of the plant or mitigation of

accidents. Therefore, the operation of the

facility in accordance with this proposed

amendment would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Jocelyn A.

Mitchell, Acting

Carolina Power & Light Company, et

al., Docket No. 50-325, Brunswick

Steam Electric Plant, Unit 1, Brunswick

County, North Carolina

Date of amendment request: April 8,

1996

Description of amendment request:

The licensee has proposed to revise the

Technical Specifications (TS) to include

the following changes: 1. The Minimum

Critical Power Ratio (MCPR) Safety

Limit specified in TS 2.1.2 from 1.07 to

1.09 for Unit 1 Cycle 11 operation; TS

5.3.1 to reflect the new fuel type (GE13)

that will be inserted during Unit 1

Refueling Outage 10; 2. The acceptable

range of sodium pentaborate

concentration for the standby liquid

control system shown in TS Figure

3.1.5-1 to reflect changes to poison

material concentration needed to

achieve reactor shutdown based on the

new GE13 fuel type.

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.

Proposed Change 1

The proposed amendment will allow the

loading and use of GE13 fuel assemblies in

the Brunswick Unit 1 reactor core. The use

of GE13 fuel assemblies requires that the

safety limit minimum critical power ratio

value also be revised. The safety limit

minimum critical power ratio is established

to maintain fuel cladding integrity during

operational transients. The GE13 fuel

assembly design has been analyzed using

methods that have been previously approved

by the Nuclear Regulatory Commission and

documented in General Electric Nuclear

Energy’s reload licensing methodology

Topical Report (NEDE-24011-P-A-11,

‘‘General Electric Standard Application for

Reactor Fuel (GESTAR II)’’ dated November

1995).

The proposed revision of the safety limit

minimum critical power ratio does not alter

any plant safety-related equipment, safety

function, or plant operations that could

change the probability of an accident. The

change does not affect the design, materials,

or construction standards applicable to the

fuel bundles in a manner that could change

the probability of an accident.

A methodology that has been previously

reviewed and accepted by the Nuclear

Regulatory Commission was used to derive

both the existing and updated safety limit

minimum critical power ratio value. The

same methodology and criteria have been

applied to derive the existing safety limit

minimum critical power ratio of 1.07 as that

used to derive the updated safety limit

minimum critical power ratio value of 1.09.

The updated safety limit minimum critical

power ratio assures that fuel cladding

protection equivalent to that provided with

the existing safety limit minimum critical

power ratio value is maintained. This ensures

that the consequences of previously

evaluated accidents are not significantly

increased.

Proposed Change 2

The standby liquid control system provides

a means of reactivity control that is

independent of the normal reactivity control

system. The standby liquid control system

must be capable of assuring that the reactor

core can be placed in a subcritical condition

at any time during reactor core life. Technical

Specification Figure 3.1.5-1 specifies the

acceptable range of concentrations and

volumes for sodium pentaborate solution

used as a neutron absorber (i.e., for reactivity

control). The portion of the sodium

pentaborate concentration range shown in

Technical Specification Figure 3.1.5-1

applicable to the lower range of tank volumes

is being revised to increase the required

concentration of sodium pentaborate

solution. This change is needed to account

for the additional shutdown reactivity

needed based on the planned use of GE13

fuel assemblies as reload fuel for the Unit 1

reactor core. Since the standby liquid control

system is independent from the normal

means of controlling reactor core reactivity

and not used to control core reactivity during

normal plant operations, the proposed

revision to the sodium pentaborate

concentration curve for the standby liquid

control system does not alter any plant

safety-related equipment, safety function, or

plant operations that could change the

probability of an accident.

The current volume-concentration range of

sodium pentaborate used in the standby

liquid control system will achieve a

sufficient concentration of boron in the

reactor vessel to ensure reactor shutdown.

Based on the increased reactivity of the new

GE13 reload fuel assemblies, the required

sodium pentaborate volume-concentration

range is being revised to ensure sufficient

neutron absorbing solution is available to

achieve reactor shutdown; therefore, the

consequences of an accident previously

evaluated are not significantly increased.

2. The proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Proposed Change 1

The GE13 fuel assembly has been designed

and complies with the acceptance criteria

contained in General Electric Nuclear

Energy’s standard application for reactor fuel

(GESTAR-II), which provides the latest

acceptance criteria for new General Electric

fuel designs. The GE13 fuel assembly

complies with GESTAR-II acceptance criteria

that have been previously reviewed and

accepted by the Nuclear Regulatory

Commission. The similarity of the GE13 fuel

design to the previously accepted GE11 fuel

design, in conjunction with the increased

critical power capability of the GE13 fuel

design, ensure that no new mode or

condition of plant operation is being

authorized by the loading and use of the
28608 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

GE13 fuel type. The proposed revision of the

safety limit minimum critical power ratio

from 1.07 to 1.09 does not modify any plant

controls or equipment that will change the

plant’s responses to any accident or transient

as given in any current analysis. Therefore,

the proposed change to allow the loading and

use of the GE13 fuel type and the revision of

the safety limit minimum critical power ratio

value from 1.07 to 1.09 will not create the

possibility for a new or different kind of

accident from any accident previously

evaluated.

Proposed Change 2

As discussed above, the standby liquid

control system provides a means of reactivity

control that is independent of the normal

reactivity control system and is capable of

assuring that the reactor core can be placed

in a subcritical condition at any time during

reactor core life. The proposed revision to the

sodium pentaborate concentration range does

not modify the standby liquid control system

or its controls, does not modify other plant

systems and equipment, and does not permit

a new or different mode of plant operation.

As such, the proposed revision to the

minimum pentaborate concentration value

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.

Proposed Change 1

As previously discussed, the GE13 fuel

assembly design has been analyzed using

methods that have been previously approved

by the Nuclear Regulatory Commission and

documented in General Electric Nuclear

Energy’s reload licensing methodology

Topical Report (NEDE-24011-P-A-11,

‘‘General Electric Standard Application for

Reactor Fuel (GESTAR II)’’ dated November

1995). The safety limit minimum critical

power ratio value is selected to maintain the

fuel cladding integrity safety limit (i.e., that

99.9 percent of all fuel rods in the core are

expected to avoid boiling transition during

operational transients). Appropriate

operating limit minimum critical power ratio

values are established, based on the safety

limit minimum critical power ratio value, to

ensure that the fuel cladding integrity safety

limit is maintained. The operating limit

minimum critical power ratio values are

incorporated in the Core Operating limits

Report as required by Technical Specification

6.9.3.1. The new GE13 safety limit minimum

critical power ratio value of 1.09 is based on

the same fuel cladding integrity safety limit

criteria [as] that for the GE11 safety limit

minimum critical power ratio value of 1.07

(i.e., that 99.9 percent of all fuel rods in the

core are expected to avoid boiling transition

during operational transients); therefore, the

proposed change does not result in a

significant reduction in the margin of safety.

Proposed Change 2

As previously stated, the purpose of the

standby liquid control is to inject a neutron

absorbing solution into the reactor in the

event that a sufficient number of control rods

cannot be inserted to maintain subcriticality.

Sufficient solution is to be injected such that

the reactor will be brought from maximum

rated power conditions to subcritical over the

entire reactor temperature range from

maximum operating to cold shutdown

conditions. General Electric methodology

establishes a fuel type dependent standby

liquid control system shutdown margin to

account for calculational uncertainties.

General Electric calculations show that an invessel concentration of 660 ppm will provide

a standby liquid control system minimum

shutdown margin in excess of the

3.2%[delta]k value required for the GE13

fuel. To achieve an in-vessel concentration of

660 ppm, the acceptable range of standby

liquid control system tank concentrations is

being revised for the lower range of tank

volumes. Thus, the proposed revision of the

standby liquid control system sodium

pentaborate volume-concentration range

ensures that there will not be a significant

reduction in the amount of available

shutdown margin and, therefore, not a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: February

27, 1996

Description of amendment request:

The proposed license amendment

would modify the Action Statement of

Technical Specification (TS) 3.7.1.1.1.

Currently, the TS action statement

requires that with the self actuation

function on one or more main steam

line code safety valves associated with

an operating loop inoperable, the

licensee must restore the inoperable

valve to operable status within 4 hours.

Otherwise, the plant must be in hot

standby within the next 6 hours and in

hot shutdown within the following 30

hours. The proposed change will allow

continued power operation at reduced

power levels with main steam safety

valves inoperable. The proposed change

is consistent with the philosophy of the

Westinghouse Standard Technical

Specifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. [The proposed change does not involve]

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change to the Action

Statement of LCO [Limiting Condition for

Operation] 3.7.1.1.1 will allow indefinite

operation at less than or equal to 75% power

in the event that the self actuation function

of no more than one safety valve per steam

generator is inoperable, and allow indefinite

operation at less than or equal to 50% power

in the event that the self actuation function

of no more than two safety valves per steam

generator is inoperable. The requirement to

reduce power will ensure that there is no

increase in the consequences of a loss of load

accident. The proposed change is consistent

with the methodology in the Westinghouse

Standard Technical Specifications. The

methodology is conservative, since the

PORVs [power operated relief valves] cannot

affect the time of reactor trip on high

pressurizer pressure. Thus, it is concluded

that the change does not increase the

consequences of any previously evaluated

accident.

The change only specifies a power

reduction in the event that the self actuation

function of steam generator safety valves is

inoperable. It does not affect the probability

of any accident. The change by itself does not

affect the likelihood of an inoperable safety

valve.

2. [The proposed change does not create]

the possibility of a new or different kind of

accident from any previously evaluated.

The change only specifies a power

reduction in the event that the self actuation

function of steam generator safety valves is

inoperable. This does not create the potential

for a new or different kind of accident. The

lower power level assures that peak steam

generator pressure and RCS [reactor coolant

system] pressure will remain below 110% of

design. This provides assurance that no

equipment failure will occur due to

overpressurization. Thus, the change does

not create the possibility for a new or

different kind of accident.

3. [The proposed change does not involve]

a significant reduction in a margin of safety.

The allowable power levels have been

selected, consistent with the Westinghouse

Standard Technical Specifications, to assure

that steam generator and RCS pressure will

remain below 110% of design. Thus, there is

no reduction in a margin of safety for

overpressure protection.

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
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28609

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: March 7,

1996

Description of amendment request:

The licensee will be replacing a locally

operated (manual) containment sump

suction isolation valve, RH-V-808A,

with a remote manually operated (motor

operated) valve, RH-MOV-808A during

the upcoming Cycle 19 refueling outage.

As a result, changes are being requested

to the Haddam Neck Plant Technical

Specifications to reflect this design

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. [The proposed change does not involve]

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed technical specification

change to Section 3/4.4.6.2 and its bases are

the replacement of the designation RH-V808A with RH-MOV-808A. There are no

changes to the requirements of this

specification and this change is therefore an

administrative change. The changes to

Section 3/4.5.1 will make the requirements

for RH-MOV-808A identical to those of RHMOV-22. RH-V-808A is being converted to a

motor operated valve (MOV). This MOV will

make the ability to establish a suction path

from the containment to the Residual Heat

Removal (RHR) System single failure proof

from the control room. Both RH-MOV-22 and

RH-MOV-808A will be opened to establish

containment sump recirculation post-loss of

coolant accident (LOCA). This will provide

added assurance that core cooling will be

maintained in the switch from injection to

containment sump recirculation following a

LOCA. The requirement for RH-MOV-808A

to be closed and its hand wheel locked can

not cause an accident. The credit for

operation of RH-MOV-808A to ensure that

the establishment of containment sump

recirculation is single failure proof is

equivalent to the current crediting of RH-V808A with the only difference being that

operation of the valve can now be performed

from the control room. Also, since both RHMOV-22 and RH-MOV-808A will be

procedurally opened during establishment of

containment sump recirculation, the

elimination of the requirement to lock open

the breaker for RH-MOV-22 will not affect the

consequences of a LOCA. The proposed

changes that reflect the conversion of RH-V808A to a MOV and the proposed changes in

how the valve is used do not increase the

consequences of a LOCA.

2. [The proposed change does not create]

the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes will require RHMOV-808A to be closed with the hand wheel

locked. This provides assurance that the

valve is in the required position. Also, RHMOV-808A will be capable of remote manual

operation during the monthly surveillance

which provides assurance that the valve can

be repositioned if necessary. The proposed

opening of RH-MOV-808A at the same time

as RH-MOV-22 is opened, provides greater

assurance that a suction path is available to

the RHR pumps as well as lowering the total

effective piping resistance from the

containment sump to the pump suction.

Therefore, the proposed changes do not

introduce the possibility of a new or different

kind of accident.

3. [The proposed change does not involve]

a significant reduction in a margin of safety.

The proposed changes make RH-MOV808A identical to RH-MOV-22 with the

exception that RH-MOV-808A will not get a

closure signal on Safety Injection Actuation.

Both RH-MOV-22 and RH-MOV-808A are

containment isolation valves in a closed

system. For closed systems, the containment

isolation requirement is that the valves be

either: a) automatic, b) locked closed, or c)

capable of remote manual operation. RHMOV-808A and RH-MOV-22 are both capable

of remote manual operation and therefore do

not need automatic closure when they are

opened as part of the technical specification

required surveillance. Therefore, the

proposed changes can not cause 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: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: March

28, 1996

Description of amendment request:

The proposed license amendment will

add an additional footnote to Limiting

Condition for Operation (LCO) 3.4.2.1

and revise an existing footnote for LCO

3.4.2.2. Currently, the footnote for LCO

3.4.2.2 requires the pressurizer code

safety valve as-found lift setting to be

within +3 percent and -1 percent of the

setpoint. The proposed change will

relax the negative as-found lift tolerance

to -3 percent. The as-left lift tolerance

will remain as plus or minus 1 percent.

The same footnote will be added to LCO

3.4.2.1.

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 change does not involve]

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change will relax the

pressurizer safety valve negative as-found lift

tolerance to -3 percent. The as-left lift

tolerance will remain as plus or minus 1

percent. This proposed technical

specification change will allow for the full

use of the plus or minus 3 percent as-found

acceptance criterion for valve testing

consistent with 1989 ASME Section XI,

Subsection IWV. The relaxing of the as-found

lift tolerance can not cause an accident. The

relaxing of the tolerance will allow the safety

valve setpoint to be closer to the Power

Operated Relief Valve (PORV) setpoint and

could result in a slightly lower pressure for

overheating events. The analysis that takes

credit for the increase in pressure to the

PORV setpoint is the Loss of Load analysis.

The minimum departure from nucleate

boiling ratio (DNBR) was reanalyzed without

taking any credit for the transient increase in

pressure. The minimum DNBR still remains

above the acceptance criterion as well as

above the limiting minimum DNBR predicted

for all Updated Final Safety Analysis Report

Chapter 15 accidents. Also, the relaxed

tolerance in conjunction with a lower safety

valve blowdown, yet still conservative,

results in a slightly higher average pressure

for a valve lift/reset cycle. This means that

pressurizer overfill will not be predicted for

the limiting transient, loss of feedwater.

Thus, the proposed relaxation of as-found lift

tolerance does not increase the probability or

consequences of the design basis accidents

previously evaluated.

2. [The proposed change does not create]

the possibility of a new or different kind of

accident from any previously analyzed.

The proposed relaxation of the lift

tolerance still requires the safety valve lift

setpoint to be above both the PORV setpoint

and the pressurizer high pressure reactor trip

setpoint. In addition, the as-left setpoint is

not being changed. The relaxed tolerance in

combination with a conservative safety valve

blowdown still will preclude the prediction

of water relief from the pressurizer. This

means that the proposed change does not

introduce the possibility of a new or different

kind of accident.
28610 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

3. [The proposed change does not involve]

a significant reduction in a margin of safety.

The proposed relaxation of the as-found lift

tolerance for valve testing is consistent with

1989 ASME Section XI, Subsection IWV. The

as-left lift tolerance will remain plus or

minus 1 percent. In addition, the design basis

analyses still meet their acceptance criteria

with the -3 percent lift tolerance. Therefore,

the proposed change can not cause 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: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: April 16,

1996

Description of amendment request:

The licensee is proposing to revise the

Technical Specifications to permit the

Haddam Neck Plant to remain in Mode

1, 2, 3, or 4 with the average water

temperature of the ultimate heat sink

(UHS) greater than 90° additional action

has been added which would require

the plant to be placed in at least Hot

Standby within 6 hours and in Cold

Shutdown within the following 30

hours upon identifying that the average

water temperature of the UHS is greater

than 95°F. In addition, the licensee is

proposing to include a new surveillance

requirement for monitoring the average

circulating water inlet temperature to be

within its limits when the average water

temperature of the UHS exceeds 89°F.

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 addition to the Action

Statement of LCO 3.7.12 of an 8 hour period

to monitor the average water temperature of

the UHS does not involve an increase in the

probability of an accident previously

evaluated. The probability of an accident

previously evaluated is not increased by a

short-term increase in the average water

temperature of the UHS. An evaluation of the

service water loads associated with the lossof-offsite power and a coincident worst case

single failure of a diesel generator to start

(resulting in the loss of two of the four

service water pumps) determined that there

is adequate margin to accomplish plant

cooldown at a service water inlet temperature

of 95°F. The recirculation phase of a LOCA

[loss-of-coolant accident] was evaluated to

verify that adequate flow would be available

to the RHR [residual heat removal] heat

exchangers. The most limiting assumptions

for the recirculation phase are offsite power

is available and one RHR heat exchanger

service water isolation valve fails to open.

The injection phase of a LOCA was evaluated

to verify that adequate flow would be

available to the CAR [containment air

recirculation] fan cooling coils. The most

limiting assumption for the injection phase is

a loss-of-offsite power. The results of these

evaluations determined that there is adequate

service water flow to accomplish plant

cooldown with average water temperature of

the UHS up to 95°F. CYAPCO [Connecticut

Yankee Atomic Power Company] also

proposes to include an additional

surveillance requirement to monitor the

average water temperature of the UHS at least

once per hour if the average water

temperature of the UHS exceeds 89°F. This

additional surveillance requirement ensures

increased operator awareness as the average

water temperature of the UHS approaches the

90°F LCO limit. Based on the above, there is

no significant increase in the consequences

of any 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 technical specification

changes do not create the possibility of a new

or different kind of accident from those

previously evaluated. The addition of an 8

hour time period to monitor the average

water temperature of the UHS increases from

6 to 14 hours the amount of time that is

allowed before the plant must proceed to Hot

Standby should the average water

temperature of the UHS increase above 90°F.

This extension of the time allowed for the

plant to be in Hot Standby does not change

the plant configuration. CYAPCO also

proposes to include an additional

surveillance requirement to monitor the

average water temperature of the UHS at least

once per hour if the average water

temperature of the UHS exceeds 89°F. This

additional surveillance requirement ensures

increased operator awareness as the average

water temperature of the UHS approaches the

90°F LCO limit.

As such, the changes do not create the

possibility of a new or different kind of

accident from those previously evaluated.

3. [The proposed change does not] involve

a significant reduction in a margin of safety.

The proposed technical specification

changes do not involve a significant

reduction in any margin of safety. The

addition of an 8 hour time period to monitor

the average water temperature of the UHS

increases from 6 to 14 hours the time

required before the plant must proceed to Hot

Standby should the average water

temperature of the UHS temperature [exceed]

90°F. An evaluation has been performed to

demonstrate that the risk significance

associated with the increased action time is

very low. In addition, safe shutdown

capability has been demonstrated for service

water inlet temperatures as high as 95°F. The

addition of a surveillance requirement to

monitor the average water temperature of the

UHS at least once per hour if the average

water temperature of the UHS exceeds 89°F

is an additional requirement, limitation, or

restriction not currently within the technical

specifications. Therefore, these changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: April 22,

1996

Description of amendment request:

The proposed amendment will allow

the use of the performance-based

containment leakage testing

requirements described in 10 CFR Part

50, Appendix J, Option B.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The changes involved in this license

amendment request revise the testing criteria

for the containment penetrations. The revised

criteria will be based on the guidance in

Regulatory Guide 1.163, ‘‘Performance-Based

Containment Leak-Test Program.’’ This

guidance allows for the use of relaxed testing

frequencies for containment penetrations that

have performed satisfactorily on a historical

basis. The Containment Leakage Rate Testing

Program considers the type of service, the

design of the penetration, and the safety

impact of the penetration in determining the
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28611

testing interval of each penetration. The NRC

Staff has reviewed the potential impact of

performance-based testing frequencies for

containment penetrations during the

development of the Option B regulation. The

NRC Staff review is documented in NUREG1493, ‘‘Performance-Based Containment

Leakage-Test Program.’’ The review

concluded that reducing the frequency of

Type A tests (Integrated Leakage Rate Tests)

from three per 10 years to one per 10 years

leads to an imperceptible increase in risk. For

Type B and C testing (Local Leakage Rate

Tests), the change in testing frequency

should not have significant impact since this

leakage contributes less than 0.1 percent of

the overall risk based on the existing

regulations. The use of Option B will allow

the extension of testing intervals with a

minimal impact on the radiological release

rates since most penetration leakage is

continually well below the specified limits.

In the accident risk evaluation, the NRC Staff

noted that the accident risk is relatively

insensitive to the containment leakage rate

because the accident risk is dominated by

accident sequences that result in failure of or

bypass of the containment. The use of a

performance-based testing program will

continue to provide assurance that the

accident analysis assumptions remain

bounding. Therefore, the proposed change

does not involve a significant increase in the

probability or consequences of an accident

previously analyzed.

Removal of the surveillance accuracy

requirement in Section 4.6.1.2.c will not

affect the probability of an accident

previously analyzed since a similar

requirement is contained in ANSI/ANS-56.8-

1994, ‘‘Containment System Leakage Testing

Requirements.’’ ANSI/ANS-56.8-1994 will be

used to develop the technical methods and

techniques for the Containment Leakage Rate

Test Program as stated in Regulatory Guide

1.163. The technical methods and techniques

in ANSI/ANS-56.8-1994 have been

determined to be acceptable to the NRC Staff.

Changes to the Administrative Section

describe the containment testing program

only and cannot increase the probability or

consequences of an accident previously

analyzed.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed license amendment does not

change the operation or equipment of the

plant. The change in the test frequency is

dependent on the establishment of a

Containment Leakage Test Program. This test

program will ensure the performance history

of each penetration is satisfactory prior to the

changing of any test frequency. Since the

performance history of the penetration will

be known, there is no possibility of the

implementation of the program creating a

new or different kind of accident than

previously analyzed. Since there is no change

to the equipment or the operation of the

plant, there is no possibility of creating a new

or different kind of accident than previously

analyzed. Therefore, the proposed change

does not create the possibility of a new or

different kind of accident from any

previously analyzed.

Removal of the surveillance accuracy

requirement in Section 4.6.1.2.c will not

create the possibility of a new or different

kind of accident from those previously

analyzed since a similar requirement is

contained in ANSI/ANS-56.8-1994,

‘‘Containment System Leakage Testing

Requirements.’’ ANSI/ANS-56.8-1994 will be

used to develop the technical methods and

techniques for the Containment Leakage Rate

Test Program as stated in Regulatory Guide

1.163. The technical methods and techniques

in ANSI/ANS-56.8-1994 have been

determined to be acceptable to the NRC staff.

Changes to the Administrative Section

describe the containment testing program

only and cannot create a different accident

from any previously analyzed.

3. Involve a significant reduction in a

margin of safety.

During the development of 10 CFR Part 50,

Appendix J, Option B, the NRC staff

determined the reduction in safety associated

with the implementation of the performancebased testing program. The results of this

review are documented in NUREG-1493. The

review concluded that reducing the

frequency of Type A tests (Integrated Leakage

Rate Tests) from three per 10 years to one per

10 years leads to an imperceptible increase

in risk. For Type B and C testing (Local

Leakage Rate Tests), the increase in testing

frequency should not have significant impact

since this leakage contributes less than 0.1

percent of the overall risk based on the

existing regulations. The use of Option B will

allow the extension of testing intervals with

a minimal impact on the radiological release

rates since most penetration leakage is

continually well below the specified limits.

In the accident risk evaluation, the NRC Staff

noted that the accident risk is relatively

insensitive to the containment leakage rate

because the accident risk is dominated by

accident sequences that result in failure of or

bypass of the containment. The use of a

performance based testing program will

continue to provide assurance that the

accident analysis assumptions remain

bounding. Therefore, this change does not

involve a significant reduction in the margin

of safety.

Removal of the surveillance accuracy

requirement in Section 4.6.1.2.c will not

involve a significant reduction in the margin

of safety since a similar requirement is

contained in ANSI/ANS-56.8-1994,

‘‘Containment System Leakage Testing

Requirements.’’ ANSI/ANS-56.8-1994 will be

used to develop the technical methods and

techniques for the Containment Leakage Rate

Test Program as stated in Regulatory Guide

1.163. The technical methods and techniques

in ANSI/ANS-56.8-1994 have been

determined to be acceptable to the NRC Staff.

Changes to the Administrative Section

describe the containment testing program

only and do not reduce the margin of safety.

Moreover, the Commission has provided

guidance concerning the application of

standards in 10 CFR 50.92 by providing

certain examples (51 FR 7751, March 6, 1986)

of amendments that are considered not likely

to involve an SHC [significant hazards

consideration]. Although the proposed

change is not enveloped by a specific

example, it has been shown that the

proposed change is not an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: February

6, 1996

Description of amendment request:

The proposed amendment would delete

the requirement to perform additional

operability testing of safety system train

components when a required

component in the redundant train

becomes inoperable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes remove the

requirement for testing which is in addition

to the normal surveillance interval. The

affected equipment is subject to periodic

surveillance testing required by the

Technical Specifications. Removing the

requirement for additional testing cannot

alter any plant operating conditions,

operating practices, equipment settings, or

equipment capabilities. Therefore, changing

an AOT [allowable outage time] or a

surveillance interval cannot increase the

probability or consequences of an accident

previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any previously evaluated?

The proposed changes remove the

requirement for testing which is in addition

to the normal surveillance interval. The

affected equipment is subject to periodic

surveillance testing required by the

Technical Specifications. Removing the

requirement for additional testing cannot

alter any plant operating conditions,

operating practices, equipment settings, or

equipment capabilities. Therefore, changing

an AOT or a surveillance interval cannot

create the possibility of a new or different
28612 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

kind of accident from any previously

evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

The proposed changes remove the

requirement for testing which is in addition

to the normal surveillance interval, in effect

extending the surveillance interval. An

excessive surveillance interval extension

could reduce the margin of safety by

reducing assurance that required equipment

will function as designed; an overly

restrictive surveillance interval could also

reduce the margin of safety by imposing

unnecessary testing wear, equipment

manipulations, and system transients on the

plant.

The existing requirements to perform

cross-train testing were based on the

operating experience available when they

were added to the TS. Typically this was

done during the initial plant licensing in

1971. The recently published Standard

Technical Specifications (NUREG 1432) do

not include cross-train testing requirements

for the Engineered Safety Features

components. It has been judged by the NRC

and by the industry, that cross-train testing

is unnecessary, and that testing at normal

surveillance intervals is adequate to assure

equipment operability. This recent judgment

is based on a much larger accumulation of

operating experience than was available at

the time Palisades was licensed. There are no

special features of the Palisades plant which

would invalidate these more recent

judgments of optimal testing requirements.

Therefore, operation of the facility in

accordance with 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: Van Wylen Library, Hope

College, Holland, Michigan 49423

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: Mark Reinhart

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request:

December 14, 1995, as supplemented by

letter dated May 16, 1996.

Description of amendment request:

The proposed amendments would

change the Technical Specifications

(TS) to improve the TS Action

Statements and Surveillance

Requirements for diesel generators in

accordance with the recommendations

and guidance in Generic Letter 93-05,

Generic Letter 94-01, NUREG-1366, and

NUREG-1431. The proposed

amendments would also incorporate

technical and administrative 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:

Criterion 1

Operation of the facilities in accordance

with the requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Improvements to the

LCOs [limiting condition for operation] and

surveillance requirements for the emergency

diesel generators do not affect their capability

to provide emergency power to plant vital

instruments and safety related equipment. In

fact, these improvements make the diesel

generators more reliable since they

significantly reduce the amount of wear and

stress due to excessive and unnecessary

testing. The proposed monthly testing of the

diesel generator continues to ensure that the

system is ready for service when needed. The

fast starts and fast loadings continue to

ensure that the timing and loading

requirements for engineered safety features

actuation are met. The proposed changes do

not affect any of the design basis accident

analyses previously evaluated. Therefore,

these proposed changes do not involve any

increase in the probability or consequences

of any accident previously evaluated. The

proposed changes are fully consistent with

the recommendations and guidance

contained in GL [Generic Letter] 93-05, GL

94-01, NUREG-1366, NUREG-1431, and are

compatible with plant operating experience.

Criterion 2

Operation of the facilities in accordance

with the requested amendments will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed changes

in fact improve the reliability of the diesel

generators by eliminating unnecessary wear

and stress. Improved reliability decreases the

failure probability which also decreases the

probability of an accident not previously

evaluated. None of the requested

amendments increase the common mode

failure probability thus would not increase

the chance of both EDG’s [emergency diesel

generators] for a particular nuclear unit being

out of service simultaneously. The proposed

changes are fully consistent with the

recommendations and guidance contained in

GL 93-05, GL 94-01, NUREG-1366, NUREG1431, and are compatible with plant

operating experience.

Criterion 3

Operation of the facilities in accordance

with the requested amendments will not

involve a significant reduction in a margin of

safety. The proposed monthly testing of the

diesel generators continues to ensure that the

system is ready for service when needed. The

fast starts and fast loadings continue to

ensure that the timing and loading

requirements for engineered safety features

actuation are met. The proposed changes

improve the reliability of the diesel

generators. Implementation of the

Maintenance Rule also ensures continued

reliability of the diesel generators. No margin

of safety is decreased as a result of these TS

changes.

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: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: April 29,

1996

Description of amendment request:

The proposed amendment relocates

several cycle specific operating

parameters from the technical

specifications to the Core Operating

Limits Report per Generic Letter 88-16.

The parameters being relocated by this

change include the variable low reactor

coolant system pressure trip (VLPT) and

the variable low reactor coolant system

pressure-temperature protective 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:

Criterion 1. Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The removal of the cycle-dependent

variable low RCS pressure-temperature

protective limits and the VLPT setpoint from

technical speciications and placing them into

the COLR has no impact on plant safety and

is considered to be administrative in nature.

The proposed change does not affect the

safety analyses, physical design, or operation

of the plant. Technical specifications will

continue to require operation within the core

protective and operational limits for each

reload cycle as calculated by the approved

reload design methodologies. The

appropriate actions required if limits are

violated will remain in the technical

specifications. The reload report presents the

results of cycle-specific evaluations of

accident analyses and transients addressed in

the ANO-1 Safety Analysis Report. The cyclespecific 10CFR50.59 evaluation of the reload
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28613

report demonstrates that changes in fuel

cycle design and the corresponding COLR do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2. Does not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed change to relocate the

variable low RCS pressure-temperature

protective limits and the VLPT setpoint from

the technical specifications to the COLR is

administrative in nature. No change to the

design configuration or method of operation

of the plant is made by this proposed change,

and therefore, no new transient initiator has

been created. Technical specifications will

continue to require operation within the

required core protective and operating limits

and appropriate actions will be taken if the

limits are exceeded. Because plant operation

will continue to be limited by the cyclespecific COLR limits that are established

using NRC-approved methodologies, these

relocations will have no impact on plant

safety.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3. Does Not Involve a Significant

Reduction in the Margin of Safety.

Existing technical specification operability

and surveillance requirements are not

reduced by the proposed change to relocate

the variable low RCS pressure-temperature

protective limits and the VLPT setpoint to

the COLR. The proposed changes are

administrative in nature and do not relate to

or modify the safety margins defined in and

maintained by the technical specifications.

The cycle-specific COLR limits for future

reload fuel cycles will continue to be

developed based on NRC approved

methodologies. Each future reload undergoes

a 10CFR50.59 evaluation to assure that

operation of the plant within the cyclespecific limits will not involve a significant

reduction in a margin of safety.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 6,

1996

Description of amendment request:

The amendment would reflect that the

name of Mississippi Power & Light

Company (MP&L) has been changed to

Entergy Mississippi, Inc. The

amendment revises Operating License

NPF-29 and Antitrust Conditions for the

Grand Gulf Nuclear Station, Unit 1

(GGNS) to (1) add the phrase ‘‘(now

renamed Entergy Mississippi, Inc.)’’

after the name of Mississippi Power &

Light Company (MP&L), (2) replace the

name of Mississippi Power & Light

Company (MP&L) by the name Entergy

Mississippi, Inc., and (3) replace a

footnote by the statement: ‘‘Amendment

—— resulted in a name change for

Mississippi Power & Light Company

(MP&L) to Entergy Mississippi,

Inc.’’.The proposed amendment

involves only a change in company

name. It does not involve any changes

to the Technical Specifications for

GGNS, or to any requirements or

limiting conditions for operation on any

equipment or any systems in the plant.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Entergy Operations, Inc. proposes to

change the current Grand Gulf Nuclear

Station Facility Operating License and

Antitrust Conditions. The specific proposed

change is to reflect that the name of one of

the companies owning Grand Gulf Nuclear

Station has legally changed from Mississippi

Power & Light Company to Entergy

Mississippi, Inc.

The Commission has provided standards

for determining whether a no significant

hazards consideration exists as stated in 10

CFR 50.92(c). A proposed amendment to an

operating license involves no significant

hazards consideration if operation of the

facility in accordance with the proposed

amendment would not: (1) involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of a

new or different kind of accident from any

accident previously evaluated; or (3) involve

a significant reduction in a margin of safety.

Entergy Operations, Inc. has evaluated the

no significant hazards consideration in its

request for this license amendment and

determined that no significant hazards

consideration results from this change. In

accordance with 10 CFR 50.91(a), Entergy

Operations, Inc. is providing the analysis of

the proposed amendment against the three

standards in 10 CFR 50.92(c). A description

of the no significant hazards consideration

determination follows:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

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, serve the same customers, and all

existing obligations and commitments will

continue unaffected.

[The proposed change does not affect any

of the existing requirements or commitments

on equipment or systems that are designed

for the safe operation of the plant. It does not

affect the design or operation of the plant.]

Therefore, the proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The administrative changes to the

Operating License [and Antitrust Condition]

requirements [to change the name of

Mississippi Power & Light] do not involve

any change in the design or operation of the

plant. The company will still own all of the

same assets, serve the same customers, and

all existing obligations and commitments

will continue unaffected.

[The proposed changes do not affect

equipment or systems that could caused an

accident at the plant.]

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change [in name] is

administrative in nature, as described above;

therefore, this change does not reduce the

level of safety imposed by any current

requirements. [The proposed changes do not

affect any equipment or systems at the plant.]

The company will still own all of the same

assets, serve the same customers, and all

existing obligations and commitments will

continue unaffected.

Therefore, the proposed changes do not

cause 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. herefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner
28614 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 8,

1996

Description of amendment request:

The amendment request would replace

the current frequency requirements in

Surveillance Requirement (SR) 3.6.1.3.5,

on the leakage rate testing for each

containment purge valve with resilient

seals, in the Technical Specifications for

Grand Gulf Nuclear Station, Unit 1

(GGNS). The proposed changes would

place these purge valves on a

performance-based leakage testing

frequency, instead of the current once

every 184 days and once within 92 days

after opening the valve.The proposed

changes do not change the limiting

conditions for operation, the required

actions for inoperability, or the other

surveillance requirements on these

primary containment isolation valves.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

In accordance with 10 CFR 50.92, Entergy

Operations, Inc. has evaluated the proposed

change to the Operating License of GGNS and

has determined that the operation of the

facility in accordance with the proposed

amendment would not involve any

significant hazards considerations. In

accordance with 10 CFR 50.91(a), Entergy

Operations, Inc. is providing the following

analysis of the proposed amendment against

the three [following] standards of 10 CFR

50.92(c):

1) The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

This change deletes the augmented testing

requirement for these containment isolation

valves and allows the surveillance intervals

to be set in accordance with the Appendix J

testing program. [Appendix J to 10 CFR Part

50 defines primary containment leakage

testing requirements for water-cooled power

reactors as GGNS and these requirements

include frequency of testing for the primary

containment isolation valves.] This change

does not affect the system function or design.

The purge valves are not an initiator of any

previously analyzed accident. Leakage rates

do not affect the probability of the occurrence

of any accident. Operating history has

demonstrated that these valves do not

degrade and cause leakage as previously

anticipated. Because these valves have been

demonstrated to be reliable, these valves can

be expected to perform the containment

isolation function as assumed in the accident

analyses.

Therefore, there is no significant increase

in the consequences of any previously

evaluated accident.

2) The proposed change would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Extending the test intervals has no

influence on, nor does it contribute in any

way to, the possibility of a new or different

kind of accident or malfunction from those

previously analyzed. No change has been

made to the design, function or method of

performing leakage testing [or to the design

and function of these valves]. Leakage

acceptance criteria have not changed. No

new accident modes are created by extending

the testing intervals. No safety-related

equipment or safety functions are altered as

a result of this change.

[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 only margin of safety that has the

potential of being impacted by the proposed

changes involves the offsite dose

consequences of postulated accidents which

are directly related to the containment

leakage rate. The proposed change does not

alter the method of performing the tests nor

does it change the leakage acceptance

criteria. Sufficient data has been collected to

demonstrate that the resilient seals do not

degrade at an accelerated rate.

[Also, the proposed change would test

these valves in accordance with the

Appendix J testing program at the plant.

Appendix J to 10 CFR Part 50 defines

primary containment leakage testing

requirements for water-cooled power reactors

as GGNS and these requirements include

frequency of testing for the primary

containment isolation valves.]

Because of this demonstrated reliability,

this change will provide sufficient

surveillance to determine an increase in the

unfiltered leakage prior to the leakage

exceeding that assumed in the accident

analysis.

Therefore, the proposed change does not

result in a significant reduction in a margin

of safety.

Based on the above evaluation, Entergy

Operation, Inc. has concluded that operation

in accordance with the proposed amendment

involves no significant hazards

considerations.

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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 9,

1996

Description of amendment request:

The amendment request would (1)

increase the safety limit minimum

critical power ratio (MCPR) for two loop

operation and single loop operation to

1.10 and 1.11, respectively, and (2) add

a General Electric topical report to the

list of documents describing the

analytical methods used to determine

the core operating limits. The proposed

changes are to Section 2.1.1, Reactor

Core Safety Limits, and Section 5.6.5,

Core Operating Limits Report (COLR),

respectively, of the Technical

Specifications (TSs).

The licensee also proposed changes to

the Bases of the TSs associated with the

above proposed 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:

Entergy Operations, Inc. proposes to

change the current Grand Gulf Nuclear

Station [GGNS] Technical Specifications. The

specific change is to modify the Minimum

Critical Power Ratio (MCPR) safety limits

reported in Technical Specification 2.1.1.2,

the list of references in Technical

Specification 5.6.5, and associated Bases

changes. The proposed change is necessary

in order to switch reload fuel vendors.

[General Electric GE11 fuel is being added to

the core in place of Siemens Power

Corporation (SPC) fuel.]

The Commission has provided standards

for determining whether no significant

hazards considerations exists as stated in 10

CFR 50.92 (c). A proposed amendment to an

operating license involves no significant

hazards 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 accident previously

evaluated; or (3) involve a significant

reduction in a margin of safety.

Entergy Operations, Inc. has evaluated the

no significant hazards consideration in its

request for this license amendment and

determined that no significant hazards

considerations result from this change. In

accordance with 10 CFR 50.91(a), Entergy

Operations, Inc. is providing the analysis of

the proposed amendment against the three

standards in 10 CFR 50.92(c). A description

of the no significant hazards consideration

determination follows:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28615

The Minimum Critical Power Ratio (MCPR)

safety limit is defined in the Bases to

Technical Specification 2.1.1 as that limit

which ‘‘ensures that during normal operation

and during Anticipated Operational

Occurrences (AOOs), at least 99.9% of the

fuel rods in the core do not experience

transition boiling.’’ The MCPR safety limit is

re-evaluated for each reload and, for GGNS

[Operating] Cycle 9, the analyses have

concluded that a two-loop MCPR safety limit

of 1.10 based on the application of the

generic GE MCPR methodology is necessary

to ensure that this acceptance criterion is

satisfied. For single-loop operation, a MCPR

safety limit of 1.11 based on the generic GE

MCPR methodology was determined to be

necessary. Core MCPR operating limits are

developed to support the Technical

Specification 3.2 requirements and ensure

these safety limits are maintained in the

event of the worst-case transient. Since the

MCPR safety limit will be maintained at all

times, operation under the proposed changes

will ensure at least 99.9% of the fuel rods in

the core do not experience transition boiling.

Therefore, The Minimum Critical Power

Ratio (MCPR) safety limit change does not

affect the probability or consequences of an

accident.

The implementation of GE’s GESTAR-II

approved methodology has no effect on the

probability or consequences of any accidents

previously evaluated. One exception to

GESTAR is that the mis-oriented and mislocated bundle events will continue to be

analyzed as accidents subject to the

acceptance criteria in the current licensing

basis. The design of the GE11 fuel bundles

is such that the bundles are not likely to be

mis-oriented or mis-located and the normal

administrative controls will be in effect for

assuring proper orientation and location.

Therefore, the probability of a fuel loading

error is not increased. This analysis ensures

that postulated dose releases will not exceed

a small fraction (10 percent) of 10CFR100

limits.

Therefore, the consequences of accidents

previously evaluated are unchanged.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The GE11 fuel to be used in [Operating]

Cycle 9 is of a design compatible with fuel

present in the core and used in the previous

cycle. Therefore, the GE11 fuel will not

create the possibility of a new or different

kind of accident. The proposed changes do

not involve any new modes of operation, any

changes to setpoints, or any plant

modifications. They introduce revised MCPR

safety limits that have been proved to be

acceptable for Cycle 9 operation. Compliance

with the applicable criterion for incipient

boiling transition continues to be ensured.

The proposed MCPR safety limits do not

result in the creation of any new precursors

to an accident.

Therefore, the proposed changes do not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The MCPR safety limits have been

evaluated to ensure that during normal

operation and during AOOs [abnormal

operating occurrences], at least 99.9% of the

fuel rods in the core do not experience

transition boiling. Therefore, the

implementation of the proposed changes in

the MCPR safety limit ensure there is no

reduction in the margin of safety.

As with the current SPC methodology,

GGNS will implement only the NRCapproved revisions to GE’s GESTAR

methodology. This GE methodology is

similar to those SPC reports currently listed

in TS 5.6.5 and it will be applied in a similar,

conservative fashion. One exception to

GESTAR is that the mis-oriented and mislocated bundle events will continue to be

analyzed as accidents subject to the

acceptance criteria in the current licensing

basis. This analysis ensures that postulated

dose releases will not exceed a small fraction

(10 percent) of 10CFR100 limits. On this

basis, the implementation of this GE

methodology 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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February

6, 1996

Description of amendment request:

The proposed change will amend the

Allowable Values of parameters in Table

3.3-4 of Waterford Steam Electric

Station, Unit 3, (Waterford 3) Technical

Specifications (TSs) to make it

consistent with the identical parameters

in Table 2.2-1 of TSs for Waterford 3.

The proposed change will add Mode 4

to the surveillance requirements of

Table 4.3-2, Item 5.c (Safety Injection

System Automatic Actuation Logic) that

was inadvertently removed. Finally, the

proposed change removes a reference to

TS 3.3.3.2 in Surveillance Requirements

TS 4.10.2.2 and 4.10.4.2 since Incore

Detectors has been removed from the

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:

The proposed changes described herein are

administrative changes necessary to correct

administrative errors. The proposed changes

will have no affect on any design basis

accidents nor will these changes affect any

material condition of the plant. Therefore,

the proposed changes will not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

The proposed changes are purely

administrative. There are no new system or

design changes associated with this proposal.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change will have no impact

on any protective boundary, safety limit, or

margin to safety. The proposed change

corrects inconsistencies in the TS and is

purely administrative in nature. Therefore,

the proposed change will not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

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

NRC Project Director: William D.

Beckner

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: May 7,

1996 (TSCR 247)

Description of amendment request:

The proposed change to the technical

specifications would adopt the

provisions of the Standard Technical

Specifications (STS), NUREG-1433, Rev.

1, which clarify surveillance

requirement applicability and allow a

maximum period of 24 hours to

complete a surveillance requirement

upon discovery that the surveillance has

been missed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not
28616 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

involve a significant increase in the

probability of occurrence or consequence of

an accident previously evaluated. The

proposed changes only affect administrative

requirements regarding the applicability and

performance of surveillances. This change

clarifies surveillance requirement

applicability and allows a maximum 24 hour

delay period for the performance of a

surveillance when it is discovered that the

surveillance has not been performed within

the required frequency, consistent with the

STS. There is minimal safety significance

associated with a delay of 24 hours in

completing the required surveillance,

particularly due to the fact that the most

probable result of any particular surveillance

performed is the successful verification of

conformance with the requirements.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed changes only affect

administrative requirements regarding the

applicability of surveillance requirements

and the performance of surveillances to allow

a maximum 24 hour delay period when it is

discovered that a surveillance has been

missed. No changes to plant equipment or

operation are affected.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in the margin

of safety since the change contained in the

proposed amendment does not change any

existing safety margins.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

GPU Nuclear Corporation, Docket No.

50-320, Three Mile Island Nuclear

Station, Unit No. 2 (TMI-2), Dauphin

County, Pennsylvania

Date of amendment request: February

16, 1995

Description of amendment request:

The proposed amendment would revise

TMI-2 Operating License No. DPR-73 by

modifying sections 4.02, 4.04, and

4.1.1.3 of the unit technical

specifications. The revisions to sections

4.02 and 4.04 would add flexibility to

the scheduling of surveillance activities

and would allow for a 24 hour period

to perform missed surveillances before

declaration of a limiting condition for

operation, respectively. These changes

would make the TMI-2 technical

specifications consistent with the

Standard Technical Specifications for

B&W Plants (NUREG-1430). The

revision to section 4.1.1.3 would allow

extension of the time interval for

surveillance of the containment airlock

doors from quarterly to annually. The

proposed changes to the TMI-2

technical specifications section 4.1.1.3

would allow a decrease in worker

exposure to radiation while maintaining

an adequate level of environmental

protection at the facility.

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:

10 CFR 50.92 provides the criteria which

the Commission uses to perform a no

significant hazards consideration. 10 CFR

50.92 states that an amendment to a facility

license involves no significant hazards if

operation of the facility in accordance with

the proposed amendment would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated, or

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated, or

3. Involve a significant reduction in a

margin of safety.

The proposed changes to the technical

specifications sections 4.02 and 4.04 are

administrative and do not involve any

physical changes to the facility. No changes

are made to operating limits or parameters,

nor to any surveillance activities. The

changes to section 4.1.1.3 extends the

interval between surveillance of the

containment airlocks; it does not change the

operability requirements, test methodology or

acceptance criteria. Based on this, GPU

Nuclear has concluded that the proposed

changes to sections 4.02 and 4.04 do not:

1. Involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The changes do not modify any

operating parameters or the release of

radioactive materials. The clarification of

maximum time extensions for surveillance is

consistent with the NRC’s Standard

Technical Specifications for Babcock and

Wilcox Plants (NUREG-1430).

2. Create the possibility of a new or

different kind of accident since these change

are administrative and no plant configuration

or operational changes are involved.

3. Involve a change in the margin of safety.

These changes are administrative in nature,

compatible with standard technical

specifications, and do not affect any safety

settings or operational limits.

GPU Nuclear has also concluded that the

proposed changes to section 4.1.1.3 do not:

1. Involve a significant increase in the

probability of occurrence of or consequences

of an accident previously evaluated. The

change to this section does not change

operating parameters, equipment operability

requirements, surveillance test methodology,

or acceptance criteria.

2. Create the possibility of a new or

different kind of accident since the change

does not affect plant equipment, plant

configuration, or plant operating parameters.

3. Involve a change in the margin of safety

since the change does not affect any

operational limits.

Based on the above analysis the licensee

concluded that the proposed changes involve

no significant safety hazards considerations

as defined by 10 CFR 50.92.

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: Government Publications

Section, State Library of Pennsylvania,

Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg,

Pennsylvania 17105

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Seymour H.

Weiss

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 1,

1996

Description of amendment request:

The proposed amendments would

change the Technical Specifications to

implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leakage-Test Program.’’

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:

[South Texas Project] STP has evaluated

the proposed Technical Specification

Amendment and determined that it does not

represent a significant hazards consideration.

Based on the criteria for defining a significant

hazards consideration established in 10 CFR

50.92, operation of STP in accordance with

the proposed amendment will not:

1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because of the

following:

10 CFR [Part] 50, Appendix J has been

amended to include provisions regarding
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28617

performance based leakage testing

requirements (Option B). Option B allows

plants with satisfactory Integrated Leak Rate

Testing (ILRT) performance history to extend

the Type A testing interval from three tests

in ten years to one test in ten years. For Type

B and Type C tests, Option B allows

extended testing interval[s] based on the leak

rate test history of each component. To be

consistent with the requirements of 10 CFR

[Part] 50, Appendix J, Option B, STP

proposes to include appropriate changes to

the Technical Specifications that incorporate

the necessary revisions associated with 10

CFR [Part] 50, Appendix J, Option B.

The proposed amendment represents the

conversion of current Technical Specification

requirements to maintain consistency with

those requirements specified by 10 CFR [Part]

50, Appendix J, Option B. The proposed

changes are consistent with the current safety

analyses. Implementation of these changes

will provide continued assurance that

specified parameters associated with

containment integrity will remain within

acceptance limits, and will not significantly

increase the probability or consequences of a

previously evaluated accident.

Some proposed changes represent minor

relaxations in current Technical

Specification requirements, but are based on

the requirements specified by Option B of 10

CFR [Part] 50, Appendix J. Changes are

consistent with the current safety analyses

and determined to represent sufficient

requirements for the assurance and reliability

of equipment assumed to operate in the

safety analyses, and provide continued

assurance that specified parameters

associated with containment integrity remain

within their acceptance limits. These changes

will not significantly increase the probability

or consequences of a previously evaluated

accident.

The systems affecting containment

integrity related to this proposed amendment

request are not assumed in any safety

analyses to initiate any accident sequence.

The probability of any accident previously

evaluated is not increased by this proposed

amendment. The proposed changes to

Technical Specification LCOs or SRs

maintain an equivalent level of reliability

and availability for all affected systems. The

proposed amendment does not increase the

consequences of any accident previously

evaluated.

There is no change to the consequences of

an accident previously evaluated because

maintaining leakage within the analyzed

limit assumed for any associated accident

analyses does not adversely affect either the

on-site or off-site dose consequences

resulting from an accident. There is no

adverse impact on the probability of accident

initiators. There is no significant increase in

the probability of any previously analyzed

accident. A plant specific risk-based analysis

of Appendix J performed for STP indicates

the containment penetration leakage dose

rate contribution to the total dose rate in

person-rem is insignificant.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

10 CFR [Part] 50, Appendix J, Option B

specifies, in part, that a Type A test which

measures both the containment system

overall integrated leakage rate at containment

pressure and system alignments assumed

during a large break LOCA [loss-of-coolant

accident], and demonstrates the capability of

primary containment to withstand an

internal pressure load, may be conducted at

an interval based on the performance of the

overall containment system. The acceptable

leakage rates are specified in the plant’s

Technical Specifications. For Type B and

Type C tests, intervals are proposed based on

the performance history of each component.

Acceptance criteria for each component is

based upon demonstration that the sum

leakage rates at design basis pressure

conditions for applicable penetrations, is

within the limit specified in the Technical

Specifications.

The proposed amendment represents the

conversion of current Technical Specification

requirements to maintain consistency with

those requirements specified in 10 CFR [Part]

50, Appendix J, Option B. The proposed

changes are consistent with the current safety

analyses. Some minor relaxations in current

Technical Specification requirements,

associated with containment integrity are

based on generic guidance provided in

Option B, NEI 94-01 and ANSI/ANS 56.8,

1994. These changes do not involve revisions

to the design of the station. Some of the

changes may involve revision in the testing

of components; however, these are in

accordance with the STP current safety

analyses and provide for appropriate testing

or surveillance that are consistent with 10

CFR [Part] 50, Appendix J, Option B. The

proposed changes will not introduce new

failure mechanisms beyond those already

considered in the current safety analyses.

The proposed amendment has been

reviewed for acceptability considering

similarity of system or component design

affecting containment integrity. No new

modes of operation are introduced by the

proposed changes. Surveillance requirements

are changed to reflect corresponding changes

associated with Option B of 10 CFR [Part] 50,

Appendix J and improvements in technique

or interval of leak rate testing performance.

The proposed changes maintain, at

minimum, the present level of operability of

any system that affects containment integrity.

The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

The associated systems that affect leak rate

integrity related to the proposed amendment,

are not assumed in any safety analysis to

initiate any accident sequence. The proposed

surveillance requirements for any affected

systems are consistent with the current

requirements specified within the Technical

Specifications and are consistent with the

requirements of Option B of 10 CFR [Part] 50,

Appendix J. The proposed surveillance

requirements maintain an equivalent level of

reliability and availability of all affected

systems and therefore, does not increase the

consequences of any previously evaluated

accident.

3) Involve a significant reduction in the

margin of safety because:

The provisions specified in Option B of 10

CFR [Part] 50 Appendix J allow changes to

Type A, Type B, and Type C test intervals

based upon the performance of past leak rate

tests. The effect of extending containment

leakage rate testing intervals has a

corresponding increase in the likelihood of

containment leakage. The degree to which

intervals can be extended is a direct function

of the potential effect to existing safety

margins and the public health and safety that

can occur due to an increased likelihood of

containment leakage. 10 CFR [Part] 50

Appendix J, Option B allows longer intervals

between leakage tests based on performance

trends but does not increase the leakage

acceptance criteria. La [maximum allowable

leakage rate] is still limited to 0.3 wt%/day.

By referencing the Containment Leakage Rate

Testing Program in LCO 3.6.1.2 ACTION, the

point at which ACTION is required is

increased from .75 La to 1.0 La. This makes

the specification consistent with the intent of

having margin between an AS-LEFT leakage

of less than or equal to .75 La and

maintaining operability with less than or

equal to 1.0 La.

Changing Appendix J test intervals from

those currently provided in the Technical

Specification to those provided in 10 CFR

[Part] 50, Appendix J, Option B, slightly

increases the risk associated with Type A,

Type B, and Type C specified accident

sequences. Historical data suggests that

increasing the Type C test interval can

slightly increase the associated risk; however,

this is compensated by the corresponding

risk reduction benefits associated with

reduction in component cycling, stress, and

wear associated with increased test intervals.

When considering the total integrated risk

which includes all analyzed accident

sequences, the risk associated with

increasing test intervals is negligible. A plant

specific risk-based analysis of Appendix J

performed for STP indicates the containment

penetration leakage dose rate contribution to

total dose rate in person-rem is insignificant.

STP proposes to revise the Technical

Specifications to be consistent with those

provisions specified in Option B of 10 CFR,

Appendix J. The proposed changes are

consistent with the STP current safety

analyses. These proposed changes do not

involve revisions to the design of the station.

The proposed changes will maintain the

same level of reliability of equipment

associated with containment integrity

assumed to operate in the safety analysis, and

provide continued assurance that specified

parameters affecting plant leak rate integrity

will remain within acceptance limits. The

proposed changes provide continued

assurance of leakage integrity of containment

without adversely affecting the public health

and safety and will not significantly reduce

existing safety margins. Plant specific riskbased analysis indicates sufficient technical

justification exists to further extend the

limits beyond those allowed by Option B.

The proposed amendment to the Technical

Specifications implements present

requirements, or the requirements in

accordance with the guidelines set forth in

Option B of 10 CFR [Part] 50, Appendix J.

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’ served as

the technical basis for Option B. STP
28618 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

performed a plant specific risk-based analysis

of containment penetration leakage dose

utilizing the same methodology used in

NUREG-1493. The analysis indicates the

containment penetration leakage dose rate

contribution to the total dose rate in personrem is insignificant. This plant specific

analysis serves to validate the applicability of

the proposed changes for STP. The proposed

changes have been approved by the NRC, are

applicable to STP, maintain necessary levels

of system or component reliability affecting

containment integrity, and do not involve a

significant reduction in the margin of safety.

The performance-based approach to

leakage rate testing concludes the impact on

public health and safety due to revised

testing intervals is negligible. The proposed

amendment will not reduce availability of

systems associated with containment

integrity when required to mitigate accident

conditions; therefore, the proposed changes

do 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 a significant relaxation of the criteria

used to establish safety limits, a significant

relaxation of the bases for limiting safety

system settings or a significant relaxation of

the bases for LCOs. Therefore, based on the

guidance provided in the Federal Register

and 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 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

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests:

December 14, 1995

Description of amendment requests:

The proposed amendments would

revise the Administrative Control

(Chapter 6) Section and other affected

Sections of the Prairie Island Technical

Specifications to generally conform with

NUREG-1431, Standard Technical

Specifications, Westinghouse Plants,

Revision 1, dated April 7, 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Operation of the Prairie Island plant in

accordance with the proposed changes does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. None of the proposed

changes involve a physical modification to

the plant, a new mode of operation or a

change to the Updated Safety Analysis

Report transient analyses. These proposed

amendments generally conform to the

guidance of NUREG-1431, Revision 1,

Section 5.0 which was previously reviewed,

accepted and issued by the NRC.

Some Section 5.0 Specifications in

NUREG-1431 were not incorporated in this

License Amendment Request. These

Specifications were not proposed because

they 1) specify requirements not currently in

the Prairie Island Technical Specifications or

otherwise committed to, 2) are addressed

elsewhere in the current Technical

Specifications, or 3) the current Technical

Specifications level of commitment is

maintained. In all these instances, the NRC

has previously reviewed and approved the

proposed level of commitment through the

issuance of the current Prairie Island

Technical specifications.

The proposed changes, in themselves, do

not reduce the level of qualification or

training such that personnel requirements

would be decreased.

In total these changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because the proposed changes, in

themselves, do not introduce a new mode of

plant operation, surveillance requirement or

involve a physical modification to the plant.

These proposed amendments generally

conform to the guidance of NUREG-1431,

Revision 1, Section 5.0 which was previously

reviewed, accepted and issued by the NRC.

Some Section 5.0 Specifications in

NUREG-1431 were not incorporated in this

License Amendment Request. These

Specifications were not proposed because

they 1) specify requirements not currently in

the Prairie Island Technical Specifications or

otherwise committed to, or 2) are addressed

elsewhere in the current Technical

Specifications. Other features are not fully

implemented but rather, the current

Technical Specification level of commitment

is maintained. In all these instances, the NRC

has previously reviewed and approved the

proposed level of commitment through the

issuance of the current Prairie Island

Technical Specifications.

In general, the proposed changes are

administrative in nature. The changes

propose to revise, delete or relocate

Specifications within the Technical

Specifications or from the Technical

Specifications to the Updated Safety Analysis

Report, plant procedures or the Operational

Quality Assurance Plan through which

adequate control is maintained. The

proposed changes do not alter the design,

function, or operation of any plant

components and therefore, no new accident

scenarios are created.

Therefore, the possibility of a new or

different kind of accident from any accident

previously evaluated would not be created

[by] these amendments.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

The proposed changes do not involve a

significant reduction in a margin of safety

because the Current Technical Specifications

requirements for safe operation of the Prairie

Island plant are maintained or increased. The

proposed changes are administrative in

nature and do not involve a physical

modification to the plant, a new mode of

operation or a change to the Updated Safety

Analysis Report transient analyses. The

proposed changes do not alter the scope of

equipment currently required to be operable

or subject to surveillance testing nor does the

proposed change affect any instrument

setpoints or equipment safety functions.

Therefore, a significant reduction in the

margin of safety would not be involved with

these amendments.

Based on the evaluation describe above,

and pursuant to 10 CFR Part 50, Section

50.91, Northern States Power Company has

determined that operation [of] the Prairie

Nuclear Generating Plant in accordance with

the proposed license amendment request

does not involve any significant hazards

considerations as defined by Nuclear

Regulatory Commission regulations in 10

CFR Part 50, Section 50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28619

NRC Project Director: Mark Reinhart

(Acting Director)

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 15, 1996

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2 to revise Technical

Specification 3.5.2, ‘‘ECCS Subsystems -

Tavg Greater Than or Equal to 350°F,’’

to change the allowed outage time for

any one safety injection pump from 72

hours to 7 days. The specific TS change

proposes to add a new footnote that

increases the allowed outage time (AOT)

for one safety injection (SI) pump from

72 hours to 7 days for performance of

non-routine, emergent maintenance and

requires review by the Plant Staff

Review Committee (PSRC), and requires

Plant Manager approval prior to

exceeding 72 hours.

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 allowed outage time (AOT)

extension does not change the operating

practices of Diablo Canyon Power Plant

(DCPP). Although the proposed change

increases the allowed time in which the

safety injection (SI) system may be out of

service for maintenance or testing, this

extended AOT will only be used in emergent

circumstances.

Increasing the AOT for the SI pumps does

not involve physical alteration of any plant

equipment and does not affect analysis

assumptions regarding functioning of

required equipment designed to mitigate the

consequences of accidents. Further, the

severity of postulated accidents and resulting

radiological effluent releases will not be

affected by the increased AOT.

Finally, the probabilistic risk assessment

determined that the increase in the core

damage probability is not considered

significant.

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.

The proposed increase to the SI pump

AOTs does not change the method by which

DCPP operates. Further, the proposed change

would not result in any physical alteration to

any plant system, and there would not be a

change in the method by 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 change does not involve

a significant reduction in a margin of safety.

There is no safety analysis impact since the

extension of the SI pump AOT interval will

have no effect on any safety limit, protection

system setpoint, or limiting condition of

operation. There is no hardware change that

would impact existing safety analysis

acceptance criteria.

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 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

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

]Date of application request: April 17,

1996

Description of amendment request:

The proposed amendment would

change Technical Specification (TS) 3/

4.3 to support a future modification to

replace existing digital portions of the

main steam and feedwater isolation

system (MSFIS) with digital processor

equipment and would authorize

revision of the FSAR to include a

description of the MSFIS modification.

The MSFIS modification is a change to

the facility, as described in the safety

analysis report, that involves an

unreviewed safety question. The

modification involves an unreviewed

safety questions because: (1) the MSFIS

design will use software which could

result in a common mode failure, (2) the

original NRC review of the MSFIS did

not evaluate 2 out of 3 coincidence

circuitry, which could introduce new

system failure modes, and (3) the MSFIS

modification utilizes manual

handswitches that could introduce new

system failure modes. The NRC will

review the modification in accordance

with 10 CFR 50.59(a)(2) in conjunction

with the review of the proposed TS

amendment.

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 addition of the MSFIS actuation logic

and relays to the TS has no adverse impact

on the probability of occurrences or the

consequences of an accident. The proposed

amendment does not change or alter the

design assumptions for the systems or

components used to mitigate the

consequences of an accident and the

methodologies used in the accident analysis

remain unchanged. The operating limits will

not be changed.

No design basis accidents will be affected

by this design change since the logic which

currently exists will continue to be

performed. Thus, the radiological

consequences will not change.

The system response time is enveloped by

the current 5 second valve stroke time. The

MSFIS response time will be less than 500

msec.

A common mode software failure could

exist if both separation groups have their

PLCs [programmable logic controllers] (3 per

train - six total) malfunction at the same time.

However, a diverse means of isolating the

feedwater lines exists given the ability of the

Main Feed Control Valves to close on a

Feedwater Isolation Signal. The MSIVs [main

steam isolation valves] do not have a diverse

means of isolating their respective steam

lines if a common mode software failure

occurs. As a result, this modification

provides a means to manually fast close the

valves at the MSFIS cabinets. The operators

will be alerted of the failure conditions of

any PLC logic channel via MCB [main control

board] annunciators and indicators. This

failure mode has a low probability of

occurrence based upon the inherent quality

of the design provided by the V&V

[verification & validation] process. Therefore,

the accident consequences are not increased

for this failure mode.

The test panel in the MSFIS cabinets has

been laid out to provide the same functions

as the existing test panel, except that PLC

status indication and coincidence logic test

functions are provided. The Emergency

Override Panel, located below the Test Panel,

provides the operator with the ability to

bypass the FWIS [feedwater isolation signal]

signal and manually fast close each MSIV as

required by the Emergency Operating

Procedures. The MSIV manual FC [fast close]

switch operation is necessary for a diverse

means of operation for software common

mode failures. The FWIS bypass switch will

allow main feedwater flow to be reestablished to each Steam Generator.
28620 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

The replacement system is functionally the

same as the current system since it performs

the same logic, receives the same inputs, and

produces the same outputs. However, the

system is more reliable and possesses triple

redundant logic. Therefore, the probability of

malfunction will not be increased.

The electrical load of the A-B PLC

equipment and existing 48 VDC [volts direct

current] actuation relays is less than that of

the existing equipment so the system will not

require any additional cooling over the

existing equipment. Proper grounding is

provided for the PLC 5 VDC and actuation

relay 48 VDC power supplies, which are

electrically isolated from each other.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The addition of the MSFIS actuation logic

and relays to the TS will not create a new

type of accident or malfunction than any

previously evaluated in the Safety Analysis

Report. The safety functions of the system are

not changed in any manner, nor is the

reliability of any structure, system or

component reduced. All design and

performance criteria continue to be met.

Since the safety functions and reliability are

not adversely affected, the proposed change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The operator’s ability to adequately

respond to an accident is not hindered by the

man-machine interface added as a result of

this modification since the operator interface

is similar to the current system and the MCB

controls will not change. The operators will

be alerted to system malfunctions through

annunciation. The current system has a

status output for each MSIV and FIV

[feedwater isolation valve] valve on the

Engineered Safety Feature Status Panel,

which will be maintained. In addition, an

isolated plant annunciator interface will

provide a MSFIS Channel Failure plant

annunciator window for both trains. Training

will be provided to the technicians,

engineers, and operators on the new features

of the system prior to installation. Therefore,

this modification does not increase the

consequential effects due to the manmachine interface.

The system is compatible with the normal

and accident environments and will be

seismically qualified in accordance with the

SNUPPS [standardized nuclear unit power

plant system] seismic spectra profile. The

equipment will be qualified for

Electromagnetic Interference concerns in

accordance with EPRI [Electric Power

Research Institute] document TR-102323-

EPRI Guideline and will meet the EPRI EMI

[electromagnetic interference] limiting

practices.

The system has the same failure mode

upon loss of power as the current system and

behaves similarly upon power restoration. A

loss of power will not result in a MSFIS

actuation.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The addition of the MSFIS actuation logic

and relays to the TS will not affect or change

a safety limit or affect plant operations. This

change will not reduce the margin of safety

assumed in the accident analysis nor reduce

any margin of safety as defined in the basis

for any TS.

The system response time for any given

valve will not exceed the required valve

stroke time. Since the MSFIS does not

contain any analog channels, no channel trip

accuracies are impacted.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037

NRC Project Director: William H.

Bateman

Virginia Electric and Power Company,

Docket Nos. 50-338 and 50-339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request: October

25, 1995

Description of amendment request:

The proposed changes would provide

an allowed outage time of 14 days for

the pressurizer power-operated relief

valve (PORV) nitrogen accumulators, as

well as provide separate action

statements for the PORV depending on

the reason for the PORV inoperability

during plant operation in power Modes

1, 2, or 3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

Specifically, operation of North Anna

Power Station in accordance with the

proposed Technical Specifications changes

will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The PORVs are assumed to mitigate the

consequences of a steam generator tube

rupture as described in the North Anna

UFSAR [Updated Final Safety Analysis

Report] as well as to limit the undesired

opening of the pressurizer safety valves for a

primary overpressure event. The proposed

action statements ensure that the steam

generator tube rupture accident analysis

requirements are met. The proposed

Technical Specification changes require the

backup nitrogen supply be available for the

PORVs to be consideredoperable and add

action statements and surveillance

requirements for the nitrogen supply

commensurate with its significance. The

proposed action statements enhance the

availability of the automatic actuation of the

PORVs by not requiring the block valves to

be closed when the backup nitrogen supplies

are inoperable. The proposed surveillance

requirements enhance the reliability of the

backup nitrogen supply to the PORVs by

verifying that there is sufficient nitrogen

pressure in the accumulators for the PORVs

to perform their design function. The

proposed Technical Specification changes do

not change any accident analyses, therefore,

the probability of any accident and its

resulting consequences are not increased.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed Technical Specification

changes do not involve any physical

modification to the plant or result in a

change in a method of operation. The backup

nitrogen supply continues to be required for

PORV operability. The proposed Technical

Specification changes provide operational

flexibility and ensure the availability of the

PORVs using the normal supply of

instrument air while the backup nitrogen

supply is being restored. This also prevents

undesirable challenges to the pressurizer

safety valves. The new surveillance

requirements verify that there is sufficient

nitrogen pressure in the accumulators for the

PORVs to perform their design functions.

3. Involve a significant reduction in a

margin of safety.

The proposed Technical Specification

changes do not affect any safety limits or

limiting safety system settings. The

availability of the PORVs will be maintained

as required in Generic Letter 90-06. The

proposed Technical Specifications will

continue to ensure that the PORVs will be

capable of performing their intended

functions.

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: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219

NRC Project Director: Eugene V.

Imbro

Wisconsin Electric Power Company,

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

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: April 24,

1996
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28621

Description of amendment request:

The proposed amendments would

revise Technical Specification (TS)

Section 15.7, ‘‘Radiological Effluent

Technical Specifications (RETS).’’

Portions of the RETS would be moved

to licensee-controlled documents

consistent with Nuclear Regulatory

Commission guidance on TS

improvements. Changes to other

sections of the TSs are also proposed

consistent with the removal of portions

of the RETS.

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 proposed amendment simplifies the

RETS and implements the recommendations

of GL 89-01 and of GL 95-10. The proposed

change relocates the operational

requirements of RETS but keeps the

programmatic controls for these requirements

in the Technical Specifications. Therefore,

the proposed changes are administrative in

nature and do not affect plant operations.

Hence, the proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated because no safetyrelated equipment, safety function, or plant

operation will be altered as a result of this

proposed change. Also, the changes are

unrelated to the initiation and mitigation of

accidents and equipment malfunctions

addressed in the Final Safety Analysis

Report.

2. Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

As stated above, the proposed action is the

relocation of the RETS procedural details to

various manuals while retaining the

administrative controls in RETS. The

relocation is consistent with the intent of the

guidance of GL 89-01 and of GL 95-10. It is

administrative and has no impact on plant

operation or safety. No safety-related

equipment, safety function, or plant

operation will be altered as a result of this

proposed change. No changes to plant

components or structures are introduced

which could create new accidents or

malfunctions not previously evaluated.

Therefore, the proposed changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because no new

accident scenario is created and no

previously evaluated accident scenario is

changed by the relocation of the procedural

details of RETS from one controlled

document to another.

3. Does the proposed amendment involve

a significant reduction in the margin of

safety?

The proposed change does not include a

change to any plant structure, system,

component, or operation. The proposed

changes do not alter the basic regulatory

requirements and do not affect any safety

analyses. The proposed change is

administrative. The procedural details of the

current RETS are relocated while the

programmatic controls consistent with

regulatory requirements, including controls

on revisions to the manuals receiving the

RETS procedural details, the Environmental

Manual (EM), Radiological Effluent Control

Program Manual (RECM), Offsite Dose

Calculation Manual (ODCM), and Process

Control Program (PCP), remain in RETS.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company,

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

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: April 29,

1996

Description of amendment request:

The proposed amendments would

revise Technical Specification (TS)

Section 15.3.14, ‘‘Fire Protection

System,’’ and Section 15.4.15, ‘‘Fire

Protection System.’’ These

specifications would be relocated to

other licensee-controlled documents in

accordance with Nuclear Regulatory

Commission generic guidance.

Additional administrative changes

consistent with the relocation are also

proposed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the

proposed Technical Specifications change

will not increase the probability or

consequences of an accident previously

evaluated.

This change request proposes to remove

certain fire protection program requirements

from the Point Beach Technical

Specifications and incorporate them into the

Final Safety Analysis Report (FSAR) and the

Fire Protection Evaluation Report (FPER). No

requirements are eliminated, modified, or deemphasized by this change. The proposed

amendment ensures that any future changes

to the fire protection program will be subject

to an appropriate evaluation in accordance

with NRC regulations to ensure that there are

no unreviewed safety questions.

Therefore, these proposed changes are

administrative in nature. There are no

proposed changes to the physical plant or the

processes which ensure the plant’s capability

to mitigate fires and achieve safe shutdown.

Therefore, there is no potential effect on the

probability or consequences of previously

evaluated accidents.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

New or different accidents can only be

created by new or different accident initiators

or sequences. Because there are no proposed

changes to the physical plant or the processes

which ensure the plant’s fire protection

capability, new or different kinds of accident

initiators will not be introduced by this

change. The proposed changes are

administrative in nature.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are

based on the design and operation of the

reactor and containment and the safety

systems that provide their protection.

Because there are no proposed changes to the

physical plant or the processes which ensure

the plant’s fire protection capability, there

will be no effect on the reactor, reactor

containment, or the safety systems which

provide their protection. Therefore, the

proposed changes will not create a reduction

in a margin of safety. The proposed changes

are administrative in nature.

Additionally, the proposed revision to

Point Beach’s operating license will not

allow Wisconsin Electric to make changes to

the approved fire protection program without

prior approval of the Nuclear Regulatory

Commission should these proposed changes

adversely affect the ability to achieve and

maintain safe shutdown in the event of a fire.

In accordance with NRC Generic Letter 86-

10, any proposed change to the approved fire

protection program requires the performance

of a 10 CFR 50.59 evaluation and a fire

hazards analysis. Should these evaluations

indicate that the ability to reach and

maintain safe shutdown has been adversely

affected, prior NRC review and approval will

be obtained prior to effecting the changes.

Thus, a significant reduction in a margin of

safety cannot occur.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Joseph P. Mann Library, 1516
28622 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: May 16,

1996. This supersedes the October 24,

1995, request published in the Federal

Register on November 27, 1995 (60 FR

58409).

Description of amendment request:

This license amendment request

proposes to revise Surveillance

Requirement 4.7.6.e.4 to reflect a

proposed design change to the output

rating, from 15kW to 5kW, of the

charcoal filter adsorber unit heater in

the pressurization system portion of the

control room emergency ventilation

system (CREVS). Surveillance

Requirements 4.7.6.c.2, 4.7.6.d, and

4.9.13.b and c, are also being revised to

reflect a proposed change to the

acceptance criteria for the testing of

carbon samples from the CREVS

charcoal adsorbers and the auxiliary/

fuel building emergency exhaust system

charcoal adsorbers. Surveillance

Requirement 4.7.7.a for the auxiliary

building portion of the auxiliary/fuel

building emergency exhaust system is

also affected by this proposed change.

However, since Surveillance

Requirement 4.7.7.a refers to

Surveillance Requirements 4.9.13.b and

c, no changes to 4.7.7.a are required.

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 design function of the filter adsorber

unit heater in the pressurization system

portion of CREVS is to reduce the relative

humidity of the air entering the charcoal

filter beds to 70% relative humidity.

Although the original design specified a

heater with a rating of 15 kW, review of the

design basis calculation for this system

indicates that only about 3.13 kW is actually

required (including applicable margins to

allow for voltage variations). The proposed

change to the CREVS heaters— output rating

from 15 kW to 5 kW will not affect the

method of operation of the system, and the

new heater capacity will still exceed filter

operational requirements and safety margin.

Neither the heater change nor the charcoal

testing protocol changes will affect system

operation or performance, nor do they affect

the probability of any event initiators. These

changes do not affect any Engineered Safety

Features actuation setpoints or accident

mitigation capabilities. Therefore, the

proposed changes will not significantly

increase the consequences of an accident or

malfunction of equipment important to safety

previously evaluated in the USAR [Updated

Safety Analysis Report].

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The requested change to the CREVS

heaters’ output rating and the changes to the

charcoal sample testing protocol will not

affect the method of operation of the systems,

and the new heater capacity will still exceed

filter operational requirements and safety

margin by a significant amount. The

proposed changes only affect the heater size

in the system and the testing criteria for the

charcoal samples. No new or different

accident scenarios, transient precursors,

failure mechanisms, or limiting single

failures will be introduced as a result of these

changes. Therefore, the possibility of a new

or different kind of accident other than those

already evaluated will not be created by this

change.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The requested change to the CREVS

heaters’ output rating will reduce the heater

output of the system, but the new heater

output will still exceed filter operational

requirements and safety margin by a

significant amount. In addition, the reduction

in heat load output from the heater will

increase the design margin between the

cooling capacity of the system air

conditioning units and the building heat

load. The new charcoal adsorber sample

laboratory testing protocol is more stringent

than the current testing practice and more

accurately demonstrates the required

performance of the adsorbers following a

design basis LOCA [loss-of-coolant accident].

Therefore, these changes will not reduce the

margin of safety of the HVAC [heating,

ventilation, and air conditioning] systems’

operation.

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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, N.W., Washington, D.C.

20037

NRC Project Director: William H.

Bateman

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.

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:

January 12, 1996, as supplemented

March 4, April 3 and April 10, 1996.

Brief description of amendments: The

amendments revise the Technical

Specification 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
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28623

changes to implement Option B for the

Type B and Type C tests were requested

by the licensee at this time.

Date of issuance: May 13, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 144 and 138

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

the Technical Specifications.

Date of initial notice in Federal

Register: January 21, 1996 (61 FR 3498);

and April 10, 1996 (61 FR 15988) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated May 13, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

March 5, 1996

Brief description of amendments:

These amendments delete the

requirement to perform a pressurizer

heater surveillance test and change the

requirement for containment visual

inspection to prevent sump clogging.

These changes are in accordance with

selected line items from NRC Generic

Letter 93-05, ‘‘Line-Item Technical

Specification Improvements to Reduce

Surveillance Requirements for Testing

During Power Operation.’’

Date of issuance: May 13, 1996

Effective date: May 13, 1996

Amendment Nos. 184 and 178Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR15989)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 13, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50-498, South Texas Project, Unit

1, Matagorda County, Texas

Date of amendment request: January

22, 1996, as supplemented by letter

dated April 18, 1996.

Brief description of amendment: The

amendment modified the steam

generator tube plugging criteria in

Technical Specification 3/4.4.5, Steam

Generators, and the associated Bases, to

allow the implementation of alternate

steam generator tube plugging criteria

for the tube-to-tubesheet joints (known

in the industry as F*) for Unit 1.

Date of issuance: May 14,

1996Effective date: May 14, 1996

Amendment No.: 82

Facility Operating License No. NPF76: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7553) The additional information

contained in the supplemental letter

dated April 18, 1996, was clarifying in

nature and thus, within the scope of the

initial notice and did not affect the

staff’s proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated May 14, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Rochester Gas and Electric

Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

February 9, 1996, as

supplementedMarch 15, 1996, and April

22, 1996.

Brief description of amendment: The

amendment revised the Administrative

Controls Section 5.6.6 of the Ginna

Technical Specifications to incorporate

a reference to the methodology for

determining pressure/temperature and

low-temperature overpressure

protection limits.

Date of issuance: May 23, 1996

Effective date: May 23, 1996

Amendment No.: 64

Facility Operating License No. DPR18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7557) The March 15, 1996, and April 22,

1996, 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 May 23, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

Rochester Gas and Electric

Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

February 9, 1996

Brief description of amendment: This

amendment changes the setpoints for

the steam generator water level-high

feedwater isolation function.Date of

issuance: May 20, 1996

Effective date: May 20, 1996

Amendment No.: 63

Facility Operating License No. DPR18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7558) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

May 20, 1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

Saxton Nuclear Experimental

Corporation (SNEC), Docket No. 50-146,

Saxton Nuclear Reactor Facility (SNEF)

Date of application for amendment:

November 21, 1995, as supplemented on

March 13, 1996.

Brief description of amendment: The

amendment adds GPU Nuclear

Corporation as a licensee for the SNEF

along with SNEC and transfers all

management-related responsibilities for

the SNEF from SNEC to GPU Nuclear

Corporation.

Date of issuance: May 10, 1996

Effective date: May 10, 1996

Amendment No.: 13Amended Facility

License No. DPR-4: Amendment

changed the Technical Specifications.

Date of initial notice in Federal

Register: January 31, 1996 (61 FR 3502).

The Commission also published a notice

of consideration of transfer of control of

license pursuant to 10 CFR 50.80 on

March 19, 1996 (61 FR 11231). The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated May 10, 1996.o

significant hazards consideration

comments received: No

Local Public Document Room

Location: Saxton Community Library,

911 Church Street, Saxton,

Pennsylvania 16678
28624 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

South Carolina Electric & Gas

Company, South Carolina Public

Service Authority, Docket No. 50-395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of application for amendment:

December 8, 1995

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to: 1) add a new

surveillance requirement to 4.1.2.2, 2)

delete 3.1.2.3 and 3.1.2.4, revise 3.4.9.3

to assure that only one charging pump

is capable of injecting water into the

primary coolant whenthe reactor is in a

shutdown mode, 4) add a new

surveillance requirement to 4.4.9.3, 5)

revise the Emergency Core Cooling

Water System pump testing acceptance

criteria, and 6) revise the BASES

supporting the above changes.

Date of issuance: May 10, 1996

Effective date: 30 days after issuance

Amendment No.: 134

Facility Operating License No. NPF12: Amendment revises the TS.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1635)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 10, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

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:

December 19, 1995, as supplemented by

letters dated January 5, 1996 and May 3,

1996.

Brief description of amendments: The

amendments replace the requirements

associated with the control room

emergency ventilation system contained

in Technical Specification Section 3/

4.7.7 with requirements related to the

operation of the control room

emergency filtration/pressurization

system and the control room air

conditioning system. In addition, a onetime extension to the allowable outage

time for the control room recirculation

filtration system is included to facilitate

implementation of design modifications.

Date of issuance: May 21, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 119 and 111

Facility Operating License Nos. NPF2 and NPF-8. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1637)

The January 5, 1996 and May 3, 1996

letters provided clarifying information

that did not change the scope of the

December 19, 1995, application and

initial proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated May 21, 1996.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 No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: April 23,

1996

Brief description of amendment: The

amendment would allow steam

generator tubes to remain in service

with bands of axial degradation in the

tube sheet region, for the remainder of

Cycle 11, provided sufficient

undegraded tubing remains to satisfy

the L*-type criteria restrictions

established by the licensee.

Date of issuance: May 20, 1996

Effective date: May 20, 1996

Amendment No.: 110

Facility Operating License No. NPF-8.

The amendment revised the Technical

Specifications. Public comments

requested as to proposed no significant

hazards consideration: Yes (61 FR

19092). 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 May 30, 1996,

but indicated that if the Commission

makes a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated May 20, 1996.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

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
Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices 28625

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 July

5, 1996, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order. As required by 10

CFR 2.714, a petition for leave to

intervene shall set forth with

particularity the interest of the

petitioner in the proceeding, and how

that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

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).
28626 Federal Register / Vol. 61, No. 109 / Wednesday, June 5, 1996 / Notices

Arizona Public Service Company, et al.,

Docket Nos. STN 50-528, STN 50-529,

and STN 50-530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of application for amendments:

May 15, 1996

Brief description of amendments: The

amendment revised Surveillance

Requirement (SR) 4.5.2.d.2 in Technical

Specification 3/4 5.2 to state that the

trisodium phosphate (TSP) contained in

the storage baskets in containment is in

the form of anhydrous TSP, rather than

dodecahydrate TSP, as currently

specified.

Date of issuance: May 15, 1996

Effective date: May 15, 1996

Amendment Nos.: Unit 1 - 107; Unit

2 - 99; Unit 3 - 79

Facility Operating License Nos. NPF41, NPF-51, and NPF-74: The

amendment 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 May 15, 1996.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

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

NRC Project Director: William H.

Bateman

Dated at Rockville, Maryland, this 29th day

of May 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation

[Doc. 96–13878 Filed 6–4–96; 8:45 am]

BILLING CODE 7590–01–9

OFFICE OF THE UNITED STATES

TRADE REPRESENTATIVE

[Docket No. 301–87]

Notice of Agreement; Monitoring and

Enforcement Pursuant to Sections 301

and 306: Canadian Exports of

Softwood Lumber

AGENCY: Office of the United States

Trade Representative.

ACTION: Notice of monitoring and

determination.

SUMMARY: On May 29, 1996, the United

States and Canada entered into an

agreement on trade in softwood lumber,

with effect form April 1, 1996. This

agreement is intended to provide a

satisfactory resolution to certain acts,

policies and practices of the

Government of Canada affecting exports

to the United States of softwood lumber

that were the subject of an investigation

initiated by the United States Trade

Representative (‘‘USTR’’) under section

302(b)(1)(A) of the Trade Act of 1974

(the Trade Act) and that were found to

be unreasonable and to burden or

restrict U.S. commerce pursuant to

section 304(a) on October 4, 1991. The

USTR has determined that this

agreement will be subject to the

provisions of section 306 of the Trade

Act and that USTR will monitor

Canadian compliance with this

agreement pursuant to section 306 of the

Trade Act and will take action under

section 301(a) if Canada fails to comply

with it.

DATES: The U.S.-Canada agreement on

trade in softwood lumber was signed on

May 29, 1996.

ADDRESSES: Office of the United States

Trade Representative, 600 17th Street,

NW, Washington, D.C. 20508.

FOR FURTHER INFORMATION CONTACT:

Gordana Earp, Deputy Assistant United

States Trade Representative for

Industry, (202) 395–6160; or William

Kane, Associate General Counsel, (202)

395–6800 (for legal issues).

SUPPLEMENTARY INFORMATION: On

October 4, 1991, Canada unilaterally

terminated a Memorandum of

Understanding (MOU) dated December

30, 1986, between the United States and

Canada under which, among other

things, Canada had imposed a 15

percent export charge on certain

softwood lumber products exported to

the United States. The MOU had been

entered into to settle a pending

countervailing duty (CVD) proceeding

examining subsidies and injury with

respect to imports of Canadian softwood

lumber. As of October 4, 1991, Canada

ceased collecting export charges under

that MOU to offset possible injurious

subsidies. In response, on October 4,

1991, (a) the U.S. Department of

Commerce announced that it would

self-initiate a CVD investigation on

softwood lumber from Canada, and (b)

the USTR initiated an investigation

pursuant to section 302(b)(1)(A) of the

Trade Act (19 U.S.C. 2412(b)(1)(A)) and

pursuant to section 304(a) of the Trade

Act determined that Canada’s acts,

policies and practices regarding the

exportation of softwood lumber to the

United States were unreasonable and

burdened or restricted U.S. commerce.

56 FR 50738 (October 8, 1991) as

amended by 46 FR 58944 (November 22,

1991).

USTR further determined that action

was appropriate under section 301 of

the Trade Act to restore and maintain

the status quo ante pending issuance of

a preliminary CVD determination, and,

if warranted, to impose duties to offset

any subsidies found in the investigation.

Commerce issued its preliminary CVD

determination on March 12, 1992 and

its final affirmative CVD determination

on May 28, 1992.

Both the domestic industry and

affected Canadian parties appealed

Commerce’s final subsidy determination

to binational panels established

pursuant to Chapter 19 of the U.S.-

Canada Free Trade Agreement (FTA).

Following completion of the panel

proceedings, and a decision by an

Extraordinary Challenge Committee

(ECC) established pursuant to FTA

Article 1904.13 affirming the results of

those proceedings, Commerce—

although it expressed disagreement with

the panel’s findings—on August 16,

1994, revoked the CVD order on

softwood lumber from Canada. 59 FR

42029 (Aug. 16, 1994). USTR

subsequently terminated the action

taken under section 301. 59 FR 52846

(October 19, 1994).

In response to the decisions of the

binational panel and the ECC, the

domestic industry filed a complaint

with the United States Court of Appeals

for the District of Columbia Circuit on

September 14, 1994, challenging

Chapter 19 of the FTA. On December 15,

1994, in order to create a process that

could ultimately settle the dispute

arising from the unilateral termination

in 1991 of the MOU by Canada, and in

conjunction with the domestic

industry’s withdrawal of its challenge to

Chapter 19 of the FTA, the United States

and Canada agreed to establish a

consultative process regarding trade in

softwood lumber. The process included

the participation of the U.S.

Government, Canadian federal and

provincial governments, and where

appropriate, industries and other

interested parties in both countries.

As a result, on May 29, 1996, the

United States and Canada entered into

an agreement on trade in softwood

lumber, with effect from April 1, 1996.

During its five-year term, the agreement

will foster stable growth in the North

American softwood lumber market and

ensure fair and competitive trade for

U.S. firms and workers by addressing

the disruptive effects of unprecedented

high levels of Canadian imports

previously found by the U.S.

Department of Commerce to be

subsidized. The agreement requires