Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32359

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 26,

1995, through June 9, 1995. The last

biweekly notice was published on

Tuesday, June 6, 1995 (60 FR 29869).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, 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 21, 1995, 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
32360 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

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, 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, 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.

Commonwealth Edison Company,

Docket Nos. STN 50-456 and STN 50-

457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of amendment request: June 8,

1995, supersedes December 16, 1994,

request in its entirety, supplemented by

letters dated November 30, 1994, April

27, 1995, May 5 and May 11, 1995.

Description of amendment request:

The proposed amendment would revise

Figure 3.4-4a in the Braidwood Unit 1’s

technical specifications which provides

the nominal pressurizer power operated

relief valve set points for the lowtemperature overpressure protection

system (LTOPS). The proposed revision

would extend the applicability of Figure

3.4-4a from 5.37 effective full power

years (EFPY) to 16 EFPY (Unit 1). In

addition, the proposed amendment

removes the 638 psig administrative

limit line from the LTOPS curve,

because the appropriate instrument

uncertainties and discharge piping

pressure limits are included in the

proposed LTOPS curve. The

amendment request also proposes

administrative changes to Figure 3.4-4a

format and its associated index page.

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 new LTOPS curve will not change any

postulated accident scenarios. The revised

curve was developed using industry

standards and regulations which are

recognized as being inherently conservative.

Appropriate instrument uncertainties and

allowances have been included in the

development of the LTOPS curves. The PT

and LTOPS curves provide RCS pressure

limits to protect the Reactor Pressure Vessel

(RPV) from brittle fracture by clearly

separating the region of normal operations

from the region where the RPV is subject to

brittle fracture.

Using Regulatory Guide (RG) 1.99,

‘‘Radiation Embrittlement of Reactor Vessel

Materials,’’ Revision 2, Braidwood Unit 1

Surveillance Capsule U and Capsule X

results and the requirements of Appendix G

to 10 CFR 50, as modified by the guidance

in ASME Code Case N-514, a new LTOPS

curve was prepared. This new curve, in

conjunction with the PT Limit curves, and

the heatup and cooldown ranges provides the

required assurance that the RPV is protected

from brittle fracture.

No changes to the design of the facility

have been made, no new equipment has been

installed, and no existing equipment has

been removed or modified. This amendment

will not change any system operating modes.

The revised LTOPS curve provides assurance

that the RPV is protected from brittle

fracture.

The index page and format changes are

purely administrative in nature and are

designed to reflect the change in the duration

of applicability of Figure 3.4-4a and improve

the readability of Figure 3.4-4a. These

administrative changes will have no effect on

any equipment, system, or operating mode.

Thus, 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 use of the new LTOPS curve does not

change any postulated accident scenarios.

The new LTOPS curve was generated using

Braidwood capsule surveillance data and an

approved, conservative methodology. No

new equipment will be installed, and no

existing equipment will be modified. No new

system interfaces are created, and no existing

system interfaces are modified. The new

LTOPS curve provides assurance that the

RPV is protected from brittle fracture.

No new accident or malfunction

mechanism is introduced by this

amendment.

The index page and format changes are

purely administrative in nature and are

designed to reflect the change in the duration

of applicability of Figure 3.4-4a, and improve

the readability of Figure 3.4-4a. These

administrative changes will have no effect on

any equipment, system, or operating mode.

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.

The new LTOPS curve was developed

using industry standards and regulations

which are recognized as being inherently

conservative. Appropriate instrument

uncertainties and allowances are included in

the development of the new LTOPS curve.

This amendment will not change the

operational characteristics or design of any

equipment or system.

All accident analysis assumptions and

conditions will continue to be met. The RPV

is adequately protected from non-ductile

failure by the revised LTOPS curve.

The index page and format changes are

purely administrative in nature and are

designed to reflect the change in the duration

of applicability of Figure 3.4-4a, and improve

the readability of Figure 3.4-4a. These

administrative changes will have no effect on

any equipment, system, or operating mode.

Thus, 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
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32361

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Wilmington Public Library,

201 S. Kankakee Street, Wilmington,

Illinois 60481

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Consumers Power Company, Docket

No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request: March 4,

1993, as revised April 14, 1993, as

supplemented April 19 and May 31,

1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

conform to the wording of the revised

10 CFR Part 20, ‘‘Standards for

Protection Against Radiation,’’ and to

reflect a separation of chemistry and

radiation protection responsibilities.

The supplemental submittals provided

additional information on the proposed

TS change in response to NRC’s request

for additional information of May 5,

1995. The original submittal was

noticed on May 12, 1993 (58 FR 28053),

as corrected June 1, 1993 (58 FR 31222).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1.Will the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change does not affect the

probability or consequences of an accident.

The proposed change is to the

ADMINISTRATIVE and RADIOLOGICAL

EFFLUENT RELEASES sections of the

facility Technical Specifications, and are

administrative in nature.

- Change ‘‘Chemistry and Radiation

Protection Supervisor’’ to ‘‘Radiation

Protection Supervisor.’’

- The change from ‘‘mR/h’’ to ‘‘mrem/h’’ is

solely a change in terminology since the

revised 10 CFR 20 does not recognize or

define the roentgen as a unit of radiation.

- The Liquid Effluents Concentration

section and the associated bases have been

revised to conform with 10 CFR 50.36(a) [10

CFR 50.36a] with effluent concentrations

limited to 10 times the limits of 10 CFR

20.1001 - 20.2402, Appendix B, Table 2,

Column 2.

- The actual instantaneous dose rate limits

of the Gaseous Effluents Dose Rate section

have not changed. However, the bases section

has. Under the former 10 CFR 20, these dose

rates correspond roughly to maximum

permissible concentration and dose(s)

received by the maximum exposed member

of the public if allowed to continue for an

entire year. These limits are used more as

instantaneous limits (dose rates above which

are not allowed to continue for more than

one hour at a time) so as to provide assurance

not to exceed 10 CFR 50, Appendix I limits.

2. Will the proposed change(s) create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

This proposed change is required by the

implementation of a new 10 CFR Part 20

requirements (except for the title change) and

are administrative in nature (sic). Neither the

material condition of the facility nor the

accident analyses are affected by this

proposed change. Therefore, the proposed

change does not create the possibility of a

different type of accident than previously

evaluated.

3. Will the proposed change involve a

significant reduction in the margin of safety?

Each limit that was affected increased the

margin of safety by making the limit more

conservative; or remained the same.

- The change of distance to ‘‘30

centimeters’’ (12 inches) is more

conservative, providing a higher degree of

protection for occupationally exposed

worker.

- The liquid effluent concentration limits

remain essentially the same. The bases have

changed to [10 CFR 50.36a] reflect 10 times

10 CFR 20.1001 - 20.2402, Appendix B, Table

2, Column 2 limits as controlled by 10 CFR

50.36(a) [10 CFR 50.36a] dose limits.

- Effluent alarm setpoints were reviewed to

determine any necessary changes and were

found to be set appropriately. No change will

be necessary.

- ‘‘The instantaneous release rate limits for

airborne releases will not be changed because

they are imposed on licensees as a control to

ensure that the licensees meet Appendix I

requirements.’’ Alarm setpoints for these

dose rate limits may change slightly due to

changes in scientific data and will be

reviewed and changed as appropriate prior to

implementation.

Therefore, the proposed change does not

involve a reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: Cynthia A.

Carpenter, Acting

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: April 12,

1995

Description of amendment request:

The amendments delete Technical

Specification 3/4.3.4, ‘‘Turbine

Overspeed Protection,’’ and its

associated Bases. The deletion of TS 3/

4.3.4 and its associated Bases provides

Duke Power Company the flexibility to

implement the manufacturer’s

recommendations for turbine steam

valve surveillance test requirements.

These test requirements will be

relocated from the TS to the Selected

Licensee Commitments (SLC) Manual.

The SLC Manual is Chapter 16 of the

Updated Final Safety Analysis Report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Relocation of the

affected TS section to the SLC Manual will

have no effect on the probability of any

accident occurring. In addition, the

consequences of an accident will not be

impacted since the above system will

continue to be utilized in the same manner

as before. No impact on the plant response

to accidents will be created.

Criterion 2

The requested amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No new accident causal

mechanisms will be created as a result of

relocating the affected TS requirements to the

SLC Manual. Plant operation will not be

affected by the proposed amendments and no

new failure modes will be created.

Criterion 3

The requested amendments will not

involve a significant reduction in a margin of

safety. No impact upon any plant safety

margins will be created. Relocation of the

affected TS requirements to the SLC Manual

in consistent with the content of the

Westinghouse RSTS [Revised Standard

Technical Specifications], as the NRC did not

require technical specification controls for

the turbine overspeed protection system in

the RSTS. The proposed amendments are

consistent with the NRC philosophy of

encouraging utilities to propose amendments

that are consistent with the content of the

RSTS.

Based upon the preceding analyses, Duke

Power Company concludes that the requested

amendments do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this
32362 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: May 18,

1995, as supplemented by letter dated

May 31, 1995.

Description of amendment request:

The proposed amendment would

change Tecnical Specification (TS)

3.6.1.2 to defer the next scheduled

containment integrated leak rate test

(ILRT) at Catawba, Unit 2, for one

outage, from the end-of-cycle (EOC) 7

refueling outage (scheduled for October

1995) to EOC-8 (scheduled for March

1997). Title 10 of the Code of Federal

Regulations, Part 50, Appendix J,

requires that three ILRTs be performed

at approximately equal intervals during

each 10-year service period at a nuclear

station. ‘‘Approximately equal

intervals’’ is defined in Catawba’s TS as

40 plus or minus 10 months. The

proposed one-time change would allow

Catawba to extend that interval to less

than or equal to 70 months.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Containment leak rate testing is not an

initiator of any accident; the proposed

interval extension does not affect reactor

operations or accident analysis, and has no

perceptible radiological consequences.

Therefore, this proposed change will not

involve a significant increase in the

probability or consequences of any

previously[]evaluated accident.

2. The proposed change will not create the

possibility of any new accident not

previously evaluated.

The proposed change does not affect

normal plant operations or configuration, nor

does it affect leak rate test methods. The test

history at Catawba (no ILRT [intergrated leak

rate test] failures) provides continued

assurance of the leak tightness of the

containment structure.

3. There is no significant reduction in a

margin of safety.

It has been documented in draft NUREG1493 that an increase in the ILRT interval

from 1 test every 3 years to 1 test every 10

years would result in an increase in

population exposure risk in the vicinity of 5

representative plants from .02% to .14%. The

proposed change included herein, an

increase from 40 [plus or minus] 10 months

to [less than or equal to] 70 months,

represents a small fraction of that already

very small increase in risk. Therefore, it may

be concluded that no significant reduction in

a margin of safety will occur.

Based on the above, no significant hazards

consideration is created by the proposed

change.

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

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: April 12,

1995

Description of amendment request:

The amendments delete Technical

Specification 3/4.3.4, ‘‘Turbine

Overspeed Protection,’’ and its

associated Bases. The deletion of TS 3/

4.3.4 and its associated Bases provides

Duke Power Company the flexibility to

implement the manufacturer’s

recommendations for turbine steam

valve surveillance test requirements.

These test requirements will be

relocated from the TS to the Selected

Licensee Commitments (SLC) Manual.

The SLC Manual is Chapter 16 of the

Updated Final Safety Analysis Report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Relocation of the

affected TS section to the SLC Manual will

have no effect on the probability of any

accident occurring. In addition, the

consequences of an accident will not be

impacted since the above system will

continue to be utilized in the same manner

as before. No impact on the plant response

to accidents will be created.

Criterion 2

The requested amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No new accident causal

mechanisms will be created as a result of

relocating the affected TS requirements to the

SLC Manual. Plant operation will not be

affected by the proposed amendments and no

new failure modes will be created.

Criterion 3

The requested amendments will not

involve a significant reduction in a margin of

safety. No impact upon any plant safety

margins will be created. Relocation of the

affected TS requirements to the SLC Manual

in consistent with the content of the

Westinghouse RSTS [Revised Standard

Technical Specifications], as the NRC did not

require technical specification controls for

the turbine overspeed protection system in

the RSTS. The proposed amendments are

consistent with the NRC philosophy of

encouraging utilities to propose amendments

that are consistent with the content of the

RSTS.

Based upon the preceding analyses, Duke

Power Company concludes that the requested

amendments do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the 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

Florida Power and Light Company, et

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

Unit No. 1, St. Lucie County, Florida

Date of amendment request: May 17,

1995

Description of amendment request:

The amendment will extend the

applicability of the current Reactor

Coolant System (RCS) Pressure/

Temperature Limits and maximum

allowed RCS heatup and cooldown rates

to 23.6 Effective Full Power Years

(EFPY) of operation. In addition,

administrative changes are proposed for
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32363

TS 3.1.2.1 (Boration Systems Flow

Paths-Shutdown) and TS 3.1.2.3

(Charging Pump-Shutdown) to clarify

the conditions for which a High

Pressure Safety Injection pump may be

used.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Pursuant to 10 CFR 50.92, a determination

may be made that a proposed license

amendment involves no significant hazards

consideration if operation of the facility in

accordance with the proposed amendment

would not: (1) involve a significant increase

in the probability or consequences of an

accident previously evaluated; or (2) create

the possibility of a new or different kind of

accident from any accident previously

evaluated; or (3) involve a significant

reduction in a margin of safety. Each

standard is discussed as follows:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The pressure-temperature (P/T) limit

curves in the Technical Specifications are

conservatively generated in accordance with

the fracture toughness requirements of 10

CFR 50 Appendix G as supplemented by the

ASME Code Section XI, Appendix G

recommendations. The RTNDT values are

based on Regulatory Guide 1.99, Revision 2,

shift prediction and attenuation formula.

Analyses of reactor vessel material

irradiation surveillance specimens are used

to verify the validity of the fluence

predictions and the P/T limit curves. Use of

these curves in conjunction with the

surveillance specimen program ensures that

the reactor coolant pressure boundary will

behave in a non-brittle manner and that the

possibility of rapidly propagating fracture is

minimized. Based on the use of plant specific

material data, analysis has demonstrated that

the current P/T limit curves will remain

conservative for up to 23.6 EFPY.

In conjunction with extending the

applicability of the existing P/T limit curves,

the low temperature overpressure protection

(LTOP) analysis for 15 EFPY is also

extended. The LTOP analysis confirms that

the current setpoints for the power-operated

relief valves (PORVs) will provide the

appropriate overpressure protection at low

Reactor Coolant System (RCS) temperatures.

Because the P/T limit curves have not

changed, the existing LTOP values have not

changed, which include the PORV setpoints,

heatup and cooldown rates, and disabling of

non-essential components.

The proposed amendment does not change

the configuration or operation of the plant,

and assurance is provided that reactor vessel

integrity will be maintained. Therefore,

operation of the facility in accordance with

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

By applying plant specific data in the

determination of critical vessel material

limits, the applicability of the existing

pressure temperature limits and LTOP

requirements can be extended. There is no

change in the configuration or operation of

the facility as a result of the proposed

amendment. The amendment does not

involve the addition of new equipment or the

modification of existing equipment, nor does

it alter the design of St. Lucie plant systems.

Therefore, operation of the facility in

accordance with the proposed amendment

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

Analysis has demonstrated that the fracture

toughness requirements of 10 CFR 50

Appendix G are satisfied and that

conservative operating restrictions are

maintained for the purpose of low

temperature overpressure protection. The P/

T limit curves will provide assurance that the

RCS pressure boundary will behave in a

ductile manner and that the probability of a

rapidly propagating fracture is minimized.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant reduction in

a margin of safety.

Based on the discussion presented above

and on the supporting Evaluation of

Proposed TS Changes, FPL has concluded

that this proposed license amendment

involves no significant hazards

consideration.

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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: May 17,

1995

Description of amendment request:

The proposed amendments will

improve consistency between the

Technical Specifications and the

improved Combustion Engineering

Standard Technical Specifications

(NUREG-1432, dated September 1992)

by incorporating changes in text and

resolving other inconsistencies

identified by the NRC and plant

operations staff.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Pursuant to 10 CFR 50.92, a determination

may be made that a proposed license

amendment involves no significant hazards

consideration if operation of the facility in

accordance with the proposed amendment

would not: (1) involve a significant increase

in the probability or consequences of an

accident previously evaluated; or (2) create

the possibility of a new or different kind of

accident from any accident previously

evaluated; or (3) involve a significant

reduction in a margin of safety. Each

standard is discussed as follows:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments consist of

administrative changes to the Technical

Specifications (TS) for St. Lucie Units 1 and

2. The amendments will implement changes

in text to improve consistency within the TS

for each unit, the improved Combustion

Engineering Standard Technical

Specifications (NUREG-1432, dated

September 1992), and the regulations. The

proposed amendments do not involve

changes to the configuration or method of

operation of plant equipment that is used to

mitigate the consequences of an accident, nor

do the changes otherwise affect the initial

conditions or conservatism assumed in any

of the plant accident analyses. Therefore,

operation of the facility in accordance with

the proposed amendments would not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed administrative revisions will

not change the physical plant or the modes

of plant operation defined in the Facility

License for each unit. The changes do not

involve the addition or modification of

equipment nor do they alter the design or

operation of plant systems. Therefore,

operation of the facility in accordance with

the proposed amendments would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not
32364 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

involve a significant reduction in a margin of

safety.

The proposed amendments are

administrative in nature and do not change

the basis for any technical specification that

is related to the establishment of, or the

preservation of, a nuclear safety margin.

Therefore, operation of the facility in

accordance with the proposed amendments

would not involve a significant reduction in

a margin of safety.

Based on the above discussion and the

supporting Evaluation of Technical

Specification changes, FPL has determined

that the proposed license amendment

involves no significant hazards

consideration.

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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J.R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Date of amendment request: May 23,

1995

Description of amendment request:

The licensee proposes to change Turkey

Point Units 3 and 4 Technical

Specifications (TS) by changing the

setpoint presentation format for the

Reactor Protection System (RPS) and

Engineered Safety Features Actuation

System (ESFAS) instrumentation

setpoints contained in Technical

Specification Tables 2.2-1 and 3.3-3.

The approved Westinghouse fivecolumn instrument setpoint

methodology currently being used to

establishing those setpoints would be

retained. The intent of the amendments

is to eliminate the need for minor

administrative license amendments to

these tables that do not impact either

the Trip Setpoints or the Safety Analysis

Limits.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

No changes to the Reactor Trip System

instrumentation setpoints, ESFAS

instrumentation setpoints, or the Turkey

Point Plant licensing basis (NRC-approved,

Westinghouse five-column setpoint

methodology, as documented in

Westinghouse topical report WCAP-12745P),

is being made. The changes proposed reduce

the level of detail in the Technical

Specifications and place that detailed

information in controlled procedures,

drawings and the Final Safety Analysis

Report. Since the setpoints and methodology

remain the same, the changes proposed by

this submittal will not increase the

probability or consequences of an accident

previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

These proposed changes remove from the

Technical Specifications a level of detail

which will be maintained in controlled

procedures and drawings. The Turkey Point

Plant licensing basis (NRC-approved,

Westinghouse five column setpoint

methodology, as documented in

Westinghouse topical report WCAP-12745P),

continues to be used to calculate the Reactor

Trip System and ESFAS setpoints. No

changes to Reactor Trip System or ESFAS

instrumentation setpoints are proposed.

Since the same methodology will be used to

determine the setpoints and no setpoints are

changed, the possibility that a new or

different kind of accident from any

previously evaluated will not be created.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The Turkey Point Plant licensing basis

(NRC-approved, Westinghouse five column

setpoint methodology, as documented in

Westinghouse topical report WCAP-12745P),

continues to be used to calculate the Reactor

Trip System and ESFAS setpoints. No

changes to the Reactor Trip System or ESFAS

instrumentation setpoints are proposed.

Since the same methodology will be used to

determine the setpoints, and no setpoints are

changed by this submittal, this change does

not involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J.R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of amendment request: June 6,

1995

Description of amendment request:

The proposed change would revise

Plant Hatch Units 1 and 2 Technical

Specification (TS) Surveillance

Requirements (SR) 3.6.4.1.3 and

3.6.4.1.4 for the secondary containment

drawdown. The revision would reduce

the SR acceptance criteria to greater

than or equal to 0.20 inch of vacuum

from greater than or equal to 0.25 inch

of vacuum. Also, the licensee proposed

to change the Bases to reflect the

proposed TS revision.

The licensee stated that the secondary

containment performs no active

function in response to either loss-ofcoolant accident or fuel handling

accident. However, its leak tightness is

required to ensure that the release of

radioactive materials from the primary

containment is restricted to those

leakage paths and associated leakage

rates assumed in the accident analysis

and that fission products entrapped

within the secondary containment

structure will be treated by the Unit 1

and Unit 2 standby gas treatment

systems prior to discharge to the

environment. This change will continue

to provide adequate margin for the

secondary containment to be

sufficiently leak tight such that the

conclusions of the accident analysis

remain valid.

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 change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The secondary containment serves

a mitigation function and therefore this

change does not increase the probability of

an accident previously evaluated. The

consequences of the previously evaluated

accidents are not affected because at the

wind conditions assumed in the accident

analysis the building will be at a negative

pressure and no exfiltration is postulated.

Furthermore, the estimated wind speed at

which exfiltration might take place (31 mph)

is not a frequent occurrence (wind speeds of

greater than 24 mph occur [less than] <0.5%

of the time based on Plant Hatch specific

meteorological data).

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

analyzed. Revising the surveillance
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32365

requirement acceptance criteria does not

physically modify the plant nor does it

modify the operation of any existing

equipment.

3. The proposed change does not involve

a significant reduction in the margin of

safety. The change in vacuum acceptance

criteria results in a slightly lower wind speed

that may result in exfiltration from the

building. However, this wind speed (31 mph)

is in the realm of wind speeds which are

infrequent at Plant Hatch. Furthermore, there

are numerous conservatisms in the existing

dose calculations including: neutral to stable

meteorological conditions, ground level

release until establishment of the required

vacuum, accident source terms at event

initiation, and no credit for plateout. The

secondary containment would be maintained

at a slight negative pressure shortly after the

Standby Gas Treatment fans are running and

the releases would be from the main stack

(well before the accident source term would

be present in the secondary containment).

Some plateout would also occur and this is

conservatively ignored. Therefore the margin

of safety is not significantly reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Herbert N.

Berkow

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: May 17,

1995

Description of amendment request:

The proposed license amendment

would revise Section 3.2 of the

Technical Specifications (TSs) for Three

Mile Island Nuclear Station, Unit 1

(TMI-1) to relocate the requirements for

volume and boron concentration of the

chemical addition system boric acid mix

tank and the reclaimed boric acid

storage tank from the TMI-1 TSs to the

TMI-1 Core Operating Limits Report.

The licensee, in its request, stated that

the proposed changes are consistent

with the intent of NRC Generic Letter

88-16.

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

involve a significant increase in the

probability of occurrence or consequences of

an accident previously evaluated. The

proposed amendment relocates chemical

addition tank volume and boron

concentration parameters from Technical

Specifications to the TMI-1 Core Operating

Limits Report. The proposed amendment

provides continued control of the values of

these parameters and assures these values are

developed using NRC-approved reload

methodologies consistent with all applicable

limits of the safety analyses addressed in the

TMI-1 [Final Safety Analysis Report] FSAR.

The Technical Specifications retain the

requirement to maintain the plant within the

appropriate bounds of these limits.

Therefore, the proposed amendment has no

effect on the probability of occurrence or

consequences of an accident previously

evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed

amendment relocates chemical addition tank

volume and boron concentration parameters

to the TMI-1 Core Operating Limits Report.

The Technical Specifications retain the

requirement to maintain the boric acid mix

tank and reclaimed boric acid storage tank

volume and boron concentration parameters

within the appropriate limits. Therefore, the

proposed amendment has no effect on the

possibility of creating a new or different kind

of accident from any accident previously

evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety. The proposed amendment provides

continued control of the boric acid mix tank

and reclaimed boric acid storage tank volume

and boron concentration parameters and

assures these values remain consistent with

all applicable limits of the safety analyses

addressed in the TMI-1 FSAR. Therefore, it

is concluded that operation of the facility in

accordance with the proposed amendment

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: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Phillip F.

McKee

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: May 24,

1995

Description of amendment request:

The proposed license amendment

would revise Table 4.1-1 of the

Technical Specifications (TSs) for Three

Mile Island Nuclear Station, Unit 1

(TMI-1) to revise the test frequency

requirement for the source range nuclear

instrumentation from 7 days before

reactor startup to 6 months before

startup.

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 TSCR would not involve

a significant increase in the probability of

occurrence or the consequences of an

accident previously evaluated.

The proposed revision to the Technical

Specifications does not involve any physical

changes to the plant, and it does not impact

the safety analysis with respect to design

basis events and assumptions. The only

change proposed is in the ‘‘Test’’ frequency

for source-range Nuclear Instrumentation by

revision of the appropriate Tech. Spec.

tables. The revised testing requirement has

no impact upon the probability of occurrence

or the consequences of any accident

previously evaluated, because no credit is

taken in the accident analyses for the source

range monitors nor are there any inputs to

the Reactor Protection System. Tech. Spec.

3.1.9.2 requires that the control rod withdraw

inhibit be operable at all times; however, it

is not affected by this change request.

Additionally, no nuclear safety equipment or

systems interface with source-range nuclear

instrumentation, and operator ability to

monitor and trend post-accident neutron

level is not affected by the proposed change.

Therefore, this change request will not

increase the probability of occurrence or the

consequences of any previously analyzed

accidents as described in the Updated [Final

Safety Analysis Report] FSAR (UFSAR).

2. Operation of the facility in accordance

with the proposed TSCR would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed revision to the TMI-1

Technical Specifications does not involve

any physical changes to the plant, and does

not impact on the safety analysis with respect

to design basis events and assumptions. The

only change proposed is in the ‘‘Test’’

frequency for Nuclear Instrumentation by

revision of the appropriate Tech. Spec.

tables. No nuclear safety equipment or
32366 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

systems interface with the source-range

nuclear instrumentation, and operator ability

to monitor and trend post-accident neutron

levels is not adversely affected by the

proposed change. In addition, the sourcerange nuclear instrument channels provide

indication to the control room, plant

computer and one of two channels provides

input to Remote Shutdown Panel B.

The 0.5% instrument drift over a six (6)

month period will not affect the ability to

operate other safety equipment; nor, will it

increase the probability of failure of the rod

withdrawal inhibit. The inhibit function is

triggered by a startup rate, and a 0.5% drift

over six (6) months will not affect the

instrument’s ability to perform the inhibit

function. Therefore, this change has no

impact upon the possibility of creating a new

or different kind of accident from any

previously evaluated in the UFSAR.

3. Operation of the facility in accordance

with the proposed TSCR would not involve

a significant reduction in a margin of safety.

The proposed revision to the TMI-1

Technical Specifications does not involve

any physical changes to the plant, and does

not impact on the safety analysis with respect

to design basis events and assumptions. The

only change proposed is in the surveillance

frequency for Nuclear Instrumentation by

revision of the appropriate Tech. Spec.

tables. Startup rate instrumentation is not

included in Technical Specifications 2.0,

‘‘Safety Limits’’; and, hence, all system

Limiting Conditions for Operation(s) remain

unchanged. Testing of the source-range

nuclear instrument channels within six (6)

months prior to a reactor startup will not

decrease the margin of safety. Hence, the

margin of safety for the plant is not

diminished by this change request.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Phillip F.

McKee

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: June 1,

1995

Description of amendment request:

The proposed license amendment

would revise Section 5.3.1.1 of the

Technical Specifications (TSs) for Three

Mile Island Nuclear Station, Unit 1

(TMI-1) to allow use of an alternate

zirconium-based cladding material

manufactured by Babcock & Wilcox

Fuel Company to test the properties of

the fuel in an operating core. Present

TSs require fuel clad material to be

either ‘‘zircaloy’’ or ‘‘ZIRLO.’’

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

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The test assemblies with the

zirconium-based claddings are mechanically

and thermal-hydraulically similar to the

remainder of the reload batch and the rest of

the core, so no failure probability is

increased, nor is any operational practice

changed which could introduce a new

initiator of an accident. The only credible

event which could occur as a result of this

demonstration is clad failure of the test fuel

rods. The number of fuel rods involved is

such a small percentage of the core inventory

that even a postulated failure of all the

demonstration fuel rods from a cause related

to the demonstration would not result in

dose consequences greater than existing

limits. A failure of the fuel rods from a cause

not related to the demonstration would not

result in consequences greater than those

which would have occurred had the

assemblies not been demonstrated

assemblies. Therefore, this change does not

increase the probability of occurrence or the

consequences of an accident previously

evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The mechanical and

thermal-hydraulic similarity of the test

assemblies to the remainder of assemblies in

the core precludes the credible possibility of

creating any new failure mode or accident

sequence. The use of the demonstration

assemblies does not involve any alterations

to plant equipment or procedures which

would introduce any new or unique

operational modes or accident precursors.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety. The demonstration assemblies meet

the same design as the remainder of

assemblies in the core. Existing reload design

and safety analysis limits are maintained,

and the FSAR analyses are bounding. No

special setpoints or other safety settings are

required as a result of the use of these two

(2) test assemblies. The assemblies will be

placed in locations which will not

experience limiting peak power conditions.

Therefore, it is concluded that operation of

the facility in accordance with the proposed

amendment does not involve a reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Phillip F.

McKee

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: April 27,

1995, as supplemented by letters dated

May 4, and May 25, 1995.

Description of amendment request:

The proposed amendment would

change the tables associated with

Technical Specification (TS) 3/4.3.3.5,

Remote Shutdown System, to eliminate

the core exit thermocouples (CETs). The

proposed amendment would also

change the tables associated with TS 3/

4.3.3.6, Accident Monitoring

Instrumentation, to require two operable

channels of CETs, where each channel

would be required to have at least two

operable CETs per core quadrant. Each

channel would also be required to have

at least four operable CETs in at least

one quadrant to support the operability

of the subcooling margin monitors. In

addition, the actions related to TS 3/

4.3.3.6 would be changed to require that

a report be submitted if one CET

channel in a quadrant is inoperable for

more than 30 days, and require a plant

shutdown if both CET channels in a

quadrant are inoperable for more than 7

days.

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 consequence of

an accident previously evaluated?

Change to Technical Specification 3.3.3.5:
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32367

Deleting the reference to the core exit

thermocouples from the Remote Shutdown

Technical Specification will not involve a

significant increase in the probability of an

accident previously evaluated because the

core exit thermocouples are not potential

accident initiators. The consequences of an

accident previously evaluated will not be

increased because the core exit

thermocouples availability is not reduced,

since adequate assurance of their operability

is provided in Technical Specification

3.3.3.6, and by the surveillance of other

indications that require the availability of the

displays that also provide the core exit

temperatures at the Auxiliary Shutdown

Panel.

Change to Technical Specification 3.3.3.6:

The proposed change reduces the number

of core exit thermocouples required per

quadrant per channel from at least 4 to at

least 2. Thus, the Actions when less than 4

thermocouples per quadrant per train are

Operable but more than 6 thermocouples per

quadrant are OPERABLE, and less than 6

thermocouples per quadrant are OPERABLE

but at least 4 thermocouples per quadrant are

OPERABLE and with the number of

OPERABLE channels less than 4

thermocouples per quadrant are being

deleted. This change does not affect the

probability of an accident. The Accident

Monitoring Instruments are not initiators of

any analyzed events. The consequence of an

accident is not affected by this change. The

requirement to have two core exit

thermocouples OPERABLE per quadrant per

channel is adequate because one OPERABLE

core exit thermocouple must be located near

the center of the core and the other

OPERABLE core exit thermocouple must be

located near the core perimeter, such that the

pair of core exit thermocouples indicate the

radial temperature gradient across their core

quadrant. The change will not alter

assumptions relative to the mitigation of an

accident or transient event. Functions

supported by the thermocouples will still be

adequately supported by the system. The

revised specification provides for at least one

quadrant per channel to have at least four

operable thermocouples to protect the

subcooling margin monitor in the event of a

single failure. The other indications used to

assess core cooling, as described in Chapter

7B of the South Texas Project Updated Final

Safety Analysis Report remain unaffected by

the proposed change. Therefore, this change

will not involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed change also affects the

allowed outage times for the thermocouples.

The existing specification allows for 31 days

in the case where there are less than four

thermocouples per quadrant per train

operable, 7 days where there are less than 6

thermocouples per quadrant, and 48 hours

where there are less than 4 thermocouples

per quadrant. The required action for each of

these cases is a plant shutdown. The

proposed specification will require a report

to the Commission after 30 days in the case

where one channel of core exit

thermocouples is inoperable, and it will

require the plant to go to HOT SHUTDOWN

if two channels are inoperable for more than

7 days. A plant shutdown with only one

channel inoperable is not warranted based on

the fact that the redundant channel remains

available to provide the necessary indication

and the passive nature of the instrumentation

(i.e., no critical automatic action).

As noted above, the core exit

thermocouples are not accident initiators;

consequently, the change in allowed outage

time does not affect the probability of an

accident. The consequences of an accident

are not significantly increased because the

changes to the allowed outage times are not

extended to allow operation of the system in

such a degraded condition that it will not

perform its function. In addition, the other

indications used to assess core cooling, as

described in Chapter 7B of the South Texas

Project Updated Final Safety Analysis Report

remain unaffected by the proposed change.

As noted above, functionality of the core exit

temperature indication is preserved by

requiring at least two thermocouples to be

operable in separate regions of the core

quadrant.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

Change to Technical Specification 3.3.3.5:

Deleting the core exit thermocouples from

the Remote Shutdown Technical

Specification will not create the possibility of

a new or different accident because there are

no automatic actuations performed by the

core exit thermocouples, nor are any different

plant configurations or different operational

procedures proposed. The existing safety

analyses are unchanged and still applicable.

Change to Technical Specification 3.3.3.6:

The proposed change reduces the number

of core exit thermocouples required per

quadrant per channel from at least 4 to at

least 2. Thus, the Actions when less than 4

thermocouples per quadrant per train are

Operable but more than 6 thermocouples per

quadrant are OPERABLE, and less than 6

thermocouples per quadrant are OPERABLE

but at least 4 thermocouples per quadrant are

OPERABLE and with the number of

OPERABLE channels less than 4

thermocouples per quadrant are being

deleted. This change will not physically alter

the plant (no new or different type of

equipment will be installed). The changes in

methods governing normal plant operation

are consistent with current safety analysis

assumptions. Therefore, the change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The change in the allowed outage time

does not alter the physical configuration of

the plant or how the plant is operated;

consequently, this change does not create the

possibility of a new or different kind of

accident.

3. Does this change involve a significant

reduction in a margin of safety?

Change to Technical Specification 3.3.3.5:

Deleting the core exit thermocouples from

the Remote Shutdown Technical

Specification does not involve a significant

reduction in the margin of safety because the

core exit thermocouples indications will still

be available at the Auxiliary Shutdown

Panel. In addition, adequate and appropriate

assurance of the operability of the core exit

thermocouples is provided in Technical

Specification 3.3.3.6 for Accident Monitoring

Instrumentation, including the changes

proposed in this letter.

Change to Technical Specification 3.3.3.6:

The proposed change reduces the number

of core exit thermocouples required per

quadrant per channel from at least 4 to at

least 2. Thus, the Actions when less than 4

thermocouples per quadrant per train are

Operable but more than 6 thermocouples per

quadrant are OPERABLE, and less than 6

thermocouples per quadrant are OPERABLE

but at least 4 thermocouples per quadrant are

OPERABLE and with the number of

OPERABLE channels less than 4

thermocouples per quadrant are being

deleted. The margin of safety is not affected

by this change. The Accident Monitoring

Instrumentation provide no automatic

actuation functions. Even though the number

of core exit thermocouples per quadrant per

channel is being reduced, the Bases

requirement to have one core exit

thermocouple located near the center of the

core and one core exit thermocouple located

near the core perimeter ensures that the pair

of core exit thermocouples indicate the radial

temperature gradient across their core

quadrant which ensures the required level of

information is available. The functions

dependent on the core exit thermocouples

are still adequately supported by the

thermocouples. The revised specification

provides for at least one quadrant per

channel to have at least four operable

thermocouples to protect the subcooling

margin monitor in the event of a single

failure. In addition, the other indications

used to assess core cooling, as described in

Chapter 7B of the South Texas Project

Updated Final Safety Analysis Report remain

unaffected by the proposed change. The

safety analysis assumptions will still be

maintained, thus, no question of safety exists.

Therefore, the change does not involve a

significant reduction in a margin of safety.

The proposed changes to the allowed

outage times have no significant impact on

the margin of safety. A plant shutdown with

only one channel inoperable is not warranted

based on the fact that the redundant channel

remains available to provide the necessary

indication and the passive nature of the

instrumentation (i.e., no critical automatic

action). Based on the small likelihood of an

accident occurring concurrent with the

station being in an ACTION statement with

regard to the thermocouples, and the small

chance that the degradation of the system in

such a situation would affect its

functionality, and the diversity provided by

other indications of core cooling, the changes

in the allowed outage times are not

considered significant.

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
32368 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

College, J. M. Hodges, Learning Center,

911 Boling Highway, Wharton, Texas

77488

Attorney for licensee: Jack R.

Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, N.W., Washington,

D.C. 20036

NRC Project Director: William D.

Beckner

Indiana Michigan Power Company,

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

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of amendment requests: March

31, 1995

Description of amendment requests:

The proposed amendments would

modify the technical specifications to

eliminate the requirement to test certain

safeguards pumps via their recirculation

flowpath. The affected pumps are the

centrifugal charging pumps, residual

heat removal pumps, motor driven

auxiliary feedwater pumps, and the

turbine driven auxiliary feedwater

pumps. The proposed amendments

would also eliminate references to

specific discharge pressures and flows

associated with these pumps and

remove footnotes associated with the

Unit 2 cycle 9-10 refueling outage which

are no longer applicable.

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:

Per 10 CFR 50.92, a proposed change does

not involve a significant hazards

consideration if the change does 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.

Criterion 1

The purpose for conducting periodic

testing of the pumps identified in this

proposed amendment is to detect gross

degradation as required by Section XI of the

ASME [American Society of Mechanical

Engineers] Code. The Cook Nuclear Plant IST

[Inservice Testing] program, which

encompasses Section XI of the ASME Code,

is the basis for the existing as well as the

proposed T/Ss. Testing the pumps utilizing

a high capacity flowpath instead of a

recirculation flow path (where applicable)

will have no impact on the ability of the

pump to perform its intended function. In

fact, it is expected that the high capacity

flowpath will provide a more accurate

assessment of the pump/systems’ conditions

and ability to meet their safety function.

The removal of specific test parameters, in

favor of referencing the Cook Nuclear Plant

IST Program, will not impact the ability of

the pumps to perform their safety related

function. IST Program parameters ensure that

the pumps under test provide the support

assumed in the plant’s safety analyses.

Therefore, based on these considerations, it

is concluded that the proposed change does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Criterion 2

The proposed change will preclude the

need to realign selected pumps to their

recirculation flowpaths for testing purposes

(where applicable). Eliminating the need for

alignment to the recirculation flowpath aids

in maximizing the pump’s availability to

perform its safety function.

As stated previously the removal of the

specific test parameters, in favor of

referencing the Cook Nuclear Plant IST

Program will not impact the ability of the

pumps to perform their intended safety

function.

Thus, it is concluded that the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

Criterion 3

As stated previously, testing of the selected

pumps utilizing a high capacity flowpath

will provide greater assurance of pump

capability and maximize pump availability.

Additionally, removing specific test

parameters in favor of referencing the Cook

Nuclear Plant IST Program will have no

impact on the ability of the pumps to perform

their intended safety function. Therefore, we

believe that the margin for safety as defined

int 10 CFR [Part] 100 has not been reduced.

Based on these considerations, it is

concluded that the 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. Although not specifically

addressed in the licensee’s analysis, the

elimination of specific discharge

pressures and flows is encompassed in

the elimination of the recirculation

testing requirement and presents no

additional significant hazards

consideration. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Cynthia A.

Carpenter, Acting

Indiana Michigan Power Company,

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

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of amendment requests: May 19,

1995

Description of amendment requests:

The proposed amendments would

modify the Technical Specification

action statement associated with the

Main Steam Safety Valves (MSSVs). The

action statement would reflect different

requirements based on operating Mode

and the power range neutron flux high

setpoint with inoperable MSSVs would

be revised in response to an issue raised

in Westinghouse Nuclear Safety

Advisory Letter 94-001.

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:

Per 10 CFR 50.92, a proposed change does

not involve a significant hazards

consideration if the change does 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.

Criterion 1

Correction of the setpoint methodology

does not represent a credible accident

initiator. The new methodology reduces the

allowable power level setpoints and is

conservative compared to the presently

evaluated setpoints. The consequences of any

previously evaluated accident are not

adversely affected by this action because the

decrease in the setpoints resulting from the

new calculational methodology will ensure

that the MSSVs are capable of relieving the

pressure at the allowable power levels. Based

on these considerations, it is concluded that

the changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

Correcting the overly restrictive action

statements of T/S 3.7.1 does not involve a

significant increase in the probability of an

accident. The proposed changes modify

existing text to more accurately reflect the

intention of the restrictions imposed by the

action statements. The changes do not create

any situation that would initiate a credible

accident sequence.

Criterion 2

The change in Table 3.7-1 reduces the

allowable power levels that can be achieved

in the event that one or more main steam

safety valve(s) is inoperable. This change is

a result of vendor guidance to correct an error

in the existing methodology used to

determine the setpoints for the power level.

Changing the methodology used to determine

the setpoints, and lowering the setpoints

themselves, do no create a new condition
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32369

that could lead to a credible accident.

Therefore, it is concluded that the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The action statements remain in effect to

perform the intended function of protecting

the plant’s secondary side when the main

steam safety valves are inoperable. They have

only been modified to correct the overly

restrictive language that specifies when, in

each MODE, specific actions must be taken.

Therefore, the proposed change does not

create a new or different type of accident.

Criterion 3

The margin of safety presently provided is

not reduced by the proposed change in the

setpoints. The change will correct the

limiting power levels that are to be

implemented when MSSVs are inoperable.

This action does not adversely affect the

margin that was previously allocated for the

ability of the MSSVs to relieve secondary

side pressure. Based on these considerations,

it is concluded that the changes do not

involve a significant reduction in a margin of

safety.

The margin of safety is also not

significantly reduced by the proposed change

to the action statements of the T/S. The

proposed revision clarifies when specific

actions are to be taken in response to

inoperable main steam safety valves. The

changes do not decrease the effectiveness of

the actions to be taken; therefore, they do not

significantly reduce any margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: Cynthia A.

Carpenter, Acting

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 16,

1995

Description of amendment request:

The proposed amendment would

modify certain requirements of the

Seabrook Station Technical

Specifications relating to containment

building penetrations during refueling

operations. One change would allow

both doors of the containment personnel

airlock (PAL) to be open during core

alterations or movement of irradiated

fuel within containment provided at

least one PAL door is capable of being

closed and a designated individual is

available outside the PAL to close the

door. Another change would allow the

use of alternate containment building

penetration closure methodologies

during refueling operations and provide

for the manual closure of a penetration

provided a designated individual is

available at the penetration.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated (10 CFR

50.92(c)(1)). The changes do not affect the

events or conditions which could result in a

fuel handling accident and do not affect any

equipment or procedures used for fuel

handling. The changes would continue to

ensure that penetrations which provide

direct access of the containment atmosphere

to outside containment are capable of

restricting a release of radioactive material to

the environment. Therefore, the changes do

not involve a significant increase in the

probability of an accident previously

evaluated.

The changes do have the potential for

increased dose at the site boundary due to a

postulated fuel handling accident. However,

the licensee’s radiological evaluations show

that the resulting offsite and control room

doses would be well within the acceptance

limits of 10 CFR Part 100 and within the

acceptance limits of GDC 19.

The Commission has provided guidance

concerning the application of standards in 10

CFR 50.92 by providing certain examples (cf.

FEDERAL REGISTER, March 6, 1986 51 FR

7751) of amendments that are considered not

likely to involve a significant hazards

consideration. These changes are similar to

example (vi) in the Federal Register notice,

in that they result in an increase in the

consequences of a previously analyzed

accident, but the results of the change are

clearly within all acceptance criteria.

B. The changes do not create the possibility

of a new or different kindof accident from

any accident previously evaluated (10 CFR

50.92(c)(2))because the changes do not affect

the events or conditions which could result

in a fuel handling accident and do not affect

any equipment or procedures used for fuel

handling. The changes do not make any

modifications to existing plant structures,

systems, or components, or otherwise affect

the manner by which the facility is operated.

C. The changes do not involve a significant

reduction in a margin ofsafety (10 CFR

50.92(c)(3)) because the increase in

calculated offsite and control room doses

resulting from a postulated fuel handling

accident are within the acceptance limits of

10 CFR Part 100 and within the acceptance

limits of GDC 19. Additionally, the changes

do not otherwise affect the manner by which

the facility is operated or involve

modifications to equipment or features which

affect the operational characteristics of the

facility.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Thomas Dignan,

Esquire, Ropes & Gray, One

International Place, Boston MA 02110-

2624.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: May 18,

1995

Description of amendment request:

The proposed amendment revises the

minimum temperature at which the

reactor vessel head bolting studs are

allowed to be placed under tension. In

addition, the proposed amendment

revises the minimum reactor vessel

metal temperature during core critical

operation, revises the minimum reactor

vessel metal temperature for pressure

tests, makes editorial changes, and

revises the bases for the applicable

section.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

changes against the criteria set forth in

10CFR50.92 and has concluded that the

changes do not involve a significant hazards

consideration (SHC). The bases for this

conclusion are that the three criteria of

10CFR50.92(c), discussed separately below,

are not compromised. The proposed changes

do not involve a SHC because the changes

would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated:

Revising the boltup temperature of the

reactor vessel head, from 86—F to 70—F,

does not decrease the margins of safety, as

required by 10CFR 50 Appendix G, against

non-ductile failure of the reactor vessel.

Therefore, the probability of occurrence of an

accident previously evaluated in the safety

analysis report (i.e., a LOCA)[loss of coolant

accident] is not increased since the revised

boltup temperature does not increase the

probability of failure of the vessel head flange

region. The reactor vessel is a passive
32370 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

component which does not initiate or play a

role in any previously evaluated accidents or

in mitigating the consequences of any

previously evaluated accidents. Therefore,

the proposed changes do not involve a

significant increase in the probability or the

consequences of any accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated:

Revising the boltup temperature of the

reactor vessel head, from 86°F to 70°F, does

not decrease the margins of safety, as

required by 10CFR 50 Appendix G, against

non-ductile failure of the reactor vessel.

Therefore, the possibility for a new or

different kind of accident than previously

evaluated (i.e., a LOCA through the vessel

flange) is not created.

3. Involve a significant reduction in a

margin of safety.

Using the proposed boltup temperature of

70°F still provides a self-imposed ‘‘margin’’

over the most limiting vessel flange region

RTNDT of 22°F (i.e., 70° - 48° = 22°). This is

a ‘‘margin’’ over and above the boltup

temperature required by Appendix G to the

1992 ASME Section XI Code, since Appendix

G would allow a boltup temperature of

48—F.

The above proposed changes to the

Limiting Condition for Operation for

tensioning the reactor vessel head studs do

not alter the configuration, normal operation,

design bases, function, mission, or

performance of the subject components.

Therefore, the proposed changes do not affect

the margin of safety inherent in the design,

analysis, function, or operation of the reactor

vessel head flange region. The proposed

changes do not alter the fuel clad barrier, fuel

integrity, reactor vessel integrity, reactor

coolant system integrity, or the containment

boundary integrity; thus the margin of safety

related to these barriers remains unchanged.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: May 24,

1995

Description of amendment request:

The proposed amendment would permit

an individual who does not have a

current senior reactor operator (SRO)

license to hold the Operations Manager

position. The position will require the

individual to have previously held an

SRO license at a boiling water reactor

(BWR). An individual serving in the

capacity of the Assistant Operations

Manager will hold a current SRO license

for Millstone Unit 1, if the Operations

Manager does not. In addition, the

proposed amendment would renumber

the applicable sections of the related

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:

NNECO has reviewed the proposed change

in accordance with 10CFR50.92 and

concluded that the change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed change does not

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The proposed change affects an

administrative control, which was based on

the guidance of ANSI N18.1-1971. ANSI

N18.1-1971 recommended that the

Operations Manager hold an SRO license.

The current guidance in Section 4.2.2 of

ANSI/ANS 3.1-1987 recommends, as one

option, that the Operations Manager have

held a license for a similar unit and the

Operations Middle Manager hold an SRO

license. While the Operations Middle

Manager position does not exist at Millstone

Unit No. 1, NNECO has created the position

of Assistant Operations Manager. The

individual in this position would meet the

requirements for, and would have

responsibilities as recommended in, ANSI/

ANS 3.1-1987 for the Operations Middle

Manager position.

Therefore, the proposed change requests an

exception to ANSI N18.1-1971 to allow use

of ANSI/ANS 3.1-1987 in a limited

circumstance. Specifically, the proposed

revision to Technical Specification 6.3.1

would require the Operations Manager to

either hold an SRO license at Millstone Unit

No. 1 or have held an SRO at a BWR.

If the Operations Manager does not hold an

SRO license at Millstone Unit No. 1, the

specification will require the Assistant

Operations Manager to hold, and continue to

hold, an SRO license. The proposed change

includes the requirement to have held a

license for a similar unit (a BWR) in

accordance with Section 4.2.2 of ANSI/ANS

3.1-1987, if the Operations Manager does not

hold an SRO license at Millstone Unit No. 1.

For those areas of knowledge that require an

SRO license, the Assistant Operations

Manager will provide the technical guidance

typically provided by the Operations

Manager.

The proposed change does not alter the

design of any system, structure, or

component, nor does it change the way plant

systems are operated. It does not reduce the

knowledge, qualifications, or skills of

licensed operators, and does not affect the

way the Operations Department is managed

by the Operations Manager. The Operations

Manager will continue to maintain the

effective performance of his personnel and

ensure the plant is operated safely and in

accordance with the requirements of the

operating license. Additionally, the control

room operators will continue to be

supervised by the licensed Shift Supervisor.

The proposed change does not detract from

the Operations Manager’s ability to perform

his primary responsibilities. In this case, by

having previously held an SRO license, the

Operations Manager has achieved the

necessary training, skills, and experience to

fully understand the operation of plant

equipment and the watch requirements for

operators. In summary, the proposed change

does not affect the ability of the Operations

Manager to provide the plant oversight

required of that position. Thus, it does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change to Technical

Specification 6.3.1 does not affect the design

or function of any plant system, structure, or

component, nor does it change the way plant

systems are operated. It does not affect the

performance of licensed operators. Operation

of the plant in conformance with technical

specifications and other license requirements

will continue to be supervised by personnel

who hold an SRO license. The proposed

change to Technical Specification 6.3.1

ensures that the Operation Manager will be

a knowledgeable and qualified individual by

requiring the individual to have held an SRO

license at a BWR. Based on the above, the

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in the

margin of safety.

The proposed change involves an

administrative control that is not related to

the margin of safety. The proposed change

does not reduce the level of knowledge or

experience required of an individual who

fills the Operations Manager position, nor

does it affect the conservative manner in

which the plant is operated. The Control

Room operators will continue to be

supervised by personnel who hold an SRO

license. Thus, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32371

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: May 26,

1995

Description of amendment request:

The proposed amendment will delete

the old limiting conditions for operation

(LCOs) and surveillance requirements

and add new LCOs, surveillance

requirements, and bases for the loss of

normal power (LNP) instrumentation

system.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed this proposed change

in accordance with 10CFR50.92 and

concluded that this change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed change does not

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The change does not increase the

probability of a loss of off-site power event

or the occurrence of any accidents which

assume loss of off-site power. This is ensured

by the LNP instrumentation system design

which uses multiple sensing relays,

redundancy, and qualified Class 1E

components, as well as conservative

operability and surveillance requirements.

Full LNP logic requires two sets of relays

to trip in one of two redundant groups. One

set monitors bus 14E and the other set

monitors bus 14F. Separate sets are provided

for loss of voltage and degraded voltage

monitoring. This design minimizes the

likelihood of an inadvertent full LNP

initiation. To maintain redundancy in the

instrumentation, two separate groups are

provided, each group being powered from an

independent DC supply. Partial LNP logic is

also provided to detect a loss of voltage on

a single emergency bus. Redundancy in the

partial LNP logic is achieved by providing an

independent logic for each emergency power

train.

The proposed technical specification

would require that the LNP instrumentation

be maintained operable except when the unit

is in cold shutdown or refueling conditions.

If redundancy in the ability to detect a loss

of voltage or degraded voltage and initiate a

full LNP is not maintained, reactor operation

would be permitted for seven days. In this

situation, both full and partial LNP (and both

emergency power sources) remain operable.

An action statement of seven days, which is

the same as the action statement duration for

an inoperable EDG [emergency diesel

generator], is justified based on continued

operability of the other LNP group.

Additionally, it allows a reasonable amount

of time to perform repairs.

The time delays and voltage setpoints

specified in Table 3.2.4 ensure that the

emergency power source starting and loading

times continue to meet the current technical

specification requirements. Also, these time

delays are long enough to preclude false trips

due to anticipated voltage transients (e.g.,

during motor starts). The relay calibration

surveillance procedure will establish

acceptance criteria for each relay to ensure

that the total times specified in Table 3.2.4

are not exceeded. The proposed surveillance

testing and calibration frequency of every

refueling outage is consistent with the

requirements in the current technical

specification.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

There are no new failure modes associated

with this change since the proposed

requirements will ensure the LNP

instrumentation system is available to

perform its safety function. Individual

voltage sensing relays, when removed from

their cases, would provide the tripped

contact configuration. The proposed

technical specification would allow relays to

be placed in the tripped condition as long as

it would not inhibit the LNP function or

cause an inadvertent initiation. Additionally,

since the design function to ensure that

adequate power is available to operate the

emergency safeguards equipment has not

changed, no new accident or accident of a

different kind is created.

3. Involve a significant reduction in the

margin of safety.

The protective boundaries are not affected

because the consequences of any design basis

accident are not changed. Since the

protective boundaries are not affected, the

safety limits are also unaffected. The

proposed change maintains the basis of the

technical specifications by ensuring that

adequate electrical power is available to

operate the emergency safeguards equipment.

By maximizing the operability of the LNP

instrumentation without requiring high risk

testing, the proposed change will improve

the margin of safety as related to availability

of electric power to safety related loads.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: January

13, 1995

Description of amendment request:

The proposed amendment would revise

the Administrative Controls Section

(6.0) of the Technical Specifications

(TS) for Hope Creek Generating Station

to reflect organizational changes and

resultant management title changes. As

indicated on the marked-up pages in

Attachment 2, PSE&G requests that: 1)

Vice President and Chief Nuclear

Officer will be replaced with Chief

Nuclear Officer and President - Nuclear

Business Unit in TS 6.1.2, 6.2.1.c,

6.5.2.4.3.g, 6.5.2.4.4.a, 6.5.2.4.4.b,

6.5.2.6, 6.6.1.b, 6.7.1.a, and 6.7.1.c. 2)

Vice President and Chief Nuclear

Officer will be replaced with Vice

President - Nuclear Operations in TS

6.5.1.8.b, and 6.5.1.9. 3) In addition,

General Manager - Quality Assurance

and Nuclear Safety will be replaced

with Director - Quality Assurance and

Nuclear Safety Review in TS 6.5.1.8.b,

6.5.1.9, 6.5.2.2, 6.5.2.4.3.g, 6.7.1.a,

6.7.1.c.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed management title changes

from Vice President and Chief Nuclear

Officer to Chief Nuclear Officer and President

- Nuclear Business Unit or Vice President -

Nuclear Operations, and from General

Manager - Quality Assurance and Nuclear

Safety to Director - Quality Assurance and

Nuclear Safety Review are administrative in

nature and do not affect assumptions

contained in the plant safety analysis, the

physical design and/or operation of the plant,

nor do they affect Technical Specifications

that preserve safety analysis assumptions.

Therefore, the proposed changes do not
32372 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Will not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The changes being proposed are purely

administrative and will not lead to material

procedure changes or to physical

modifications. Therefore, the proposed

changes do not create the possibility of a new

or different type of accident from any

accident previously evaluated.

3. Will not involve a significant reduction

in a margin of safety.

The changes being proposed are

administrative in nature and do not relate to

or modify the safety margins defined in and

maintained by the Technical Specifications.

The changes discussed herein do not reduce

the Technical Specification safety margin

since all organizational responsibilities are

being adequately implemented, and all

personnel in place are properly qualified.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: May 19,

1995 (TS 95-07)

Description of amendment request:

The proposed change would (1) modify

Surveillance Requirement (SR) 4.1.1.3 to

allow suspension of the end of life

(EOL) moderator temperature coefficient

(MTC) surveillance measurement

provided the benchmark criteria and the

Revised Prediction as documented in

the Core Operating Limits Report

(COLR) are satisfied. The SR would also

indicate that the data required for the

calculation of the Revised Prediction is

provided in the Most Negative

Temperature Coefficient Limit Report

per Specification 6.9.1.15. In addition, a

grammatical error affecting the Unit 1

SR would be corrected; (2) modify

Technical Specifications (TS) 6.9.1.14,

COLR, by adding to the list of

references: WCAP-13749-P-[A], ‘‘Safety

Evaluation Supporting the Conditional

Exemption of the Most Negative EOL

Moderator Temperature Coefficient

Measurement,’’ May 1993 (Proprietary)

(Methodology for Specification 3.1.1.3 -

Moderator Temperature Coefficient); (3)

add Specification 6.9.1.15, which would

require that the Most Negative MTC

Report be prepared at least 60 days prior

to the date the limit would become

effective and be maintained on file.

Also, the TS would require that the data

required for the determination of the

Revised Prediction of the 300 ppm/RTP

MTC per WCAP-13749-P-[A] be

included in the report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The conditional exemption of the most

negative moderator temperature coefficient

(MTC) measurement does not change the

most negative MTC surveillance requirement

(SR) and limiting condition of operation

(LCO) limits in the TSs. Since these MTC

values are unchanged, and since the basis for

the derivation of these values from the safety

analysis moderator density coefficient (MDC)

is unchanged, the constant MDC assumed for

the Updated Final Safety Analysis Report

(UFSAR) safety analyses will also remain

unchanged. Therefore, no change in the

modeling (i.e., probabilities) of the accident

analysis conditions or response is necessary

in order to implement the change to the

conditional exemption methodology. In

addition, since the constant MDC assumed in

the safety analyses is not changed by the

conditional exemption of the most negative

MTC SR measurement, the consequences of

an accident previously evaluated in the

UFSAR are not increased. The dose

predictions presented in the UFSAR for a

steam generator tube rupture remain valid

such that more severe consequences will not

occur. Additionally, since mass and energy

releases for a loss-of-coolant accident and a

steamline break are not increased as a result

of the unchanged MDC, the dose predictions

for these events presented in the UFSAR also

remain bounding.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Since the end-of-life MTC is not changed

by the conditional exemption methodology of

WCAP-13749-P, the possibility of an

accident, which is different than any already

evaluated in the UFSAR, has not been

created. No new or different failure modes

have been defined for any system or

component nor has any new limiting single

failure been identified. Conservative

assumptions for the MDC have already been

modeled in the UFSAR analyses. These

assumptions will remain valid since the

conditional exemption methodology

documented in WCAP-13749-P does not

change the safety analysis MDC nor the TS

values of the MTC.

3. Involve a significant reduction in a

margin of safety.

The conditional exemption methodology is

documented in WCAP-13749-P. This WCAP

has been evaluated (Reference: SECL 93-

117,R1) relative to the design basis, including

the TSs, and has been determined to bound

the conditions under which the

specifications permit operation. The results

as presented in the UFSAR remain bounding

since the MDC assumed in the safety

analyses and the limiting conditions for

operation and SR MTCs in the TSs remain

unchanged. Therefore, the margin of safety,

as defined in the bases to these TSs, is not

reduced.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket No.

50-328, Sequoyah Nuclear Plant, Unit 2,

Hamilton County, Tennessee

Date of amendment request: May 19,

1995 (TS 95-13)

Description of amendment request:

The proposed change would revise

License Condition 2.C.(17) to extend the

required surveillance interval to May 4,

1996, for Surveillance Requirement

4.3.2.1.3. The proposed change would

extend the Engineered Safety Features

Response Time instrument tests

required at 36-month intervals shown in

Table 3.3-3 associated with safety

injection, feedwater isolation,

containment isolation Phase A,

auxiliary feedwater pump, essential raw

cooling water system, emergency gas

treatment system, containment spray,

containment isolation Phase B, turbine

trip, 6.9-kilovolt shutdown boarddegraded voltage or loss of voltage, and

automatic switchover to containment

sump actuations. The proposed

extension will limit the interval past the

allowable extension provided by TS

4.0.2 to 4.5 months.
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32373

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c).

Operation of Sequoyah Nuclear Plant

(SQN) in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is temporary and

allows a one-time extension of Surveillance

Requirement 4.3.2.1.3 for Cycle 7 to allow

surveillance testing to coincide with the

seventh refueling outage. The proposed

surveillance interval extension will not cause

a significant reduction in system reliability

nor affect the ability of the systems to

perform their design function. Current

monitoring of plant conditions and

continuation of the surveillance testing

required during normal plant operation will

continue to be performed to ensure

conformance with TS operability

requirements. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Extending the surveillance interval for the

performance of specific testing will not create

the possibility of a new or different kind of

accidents. No changes are required to any

system configurations, plant equipment, or

analyses. Therefore, this change will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

Surveillance interval extensions will not

impact any plant safety analyses since the

assumptions used will remain unchanged.

The safety limits assumed in the accident

analyses and the design function of the

equipment required to mitigate the

consequences of any postulated accidents

will not be changed since only the

surveillance test interval is being extended.

Historical performance generally indicates a

high degree of reliability, and surveillance

testing performed during normal plant

operation will continue to be performed to

verify proper performance. Therefore, the

plant will be maintained within the analyzed

limits, and the proposed extension will not

significantly reduce the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request: May 1,

1995

Brief description of amendments: The

proposed amendment would: (1) reduce

the minimum fuel oil volume

requirement during MODES 5 and 6, for

OPERABLE emergency diesel generators

(EDG), and (2) allow continued

OPERABLE status of diesel generators

during all MODES, for 48 hours with

greater than 6-day supply of diesel fuel

for the associated diesel generator.

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?

REDUCTION IN MINIMUM DIESEL FUEL

STORED VOLUME WHILE SHUTDOWN

The first proposed change reduces the

diesel fuel oil inventory required during

plant shutdown conditions (MODES 5 and 6).

The current fuel oil inventory requirement is

the same for plant operation (MODES 1, 2, 3

and 4) and for plant shutdown. This current

inventory requirement is based upon the

seven days continuous operation of a diesel

generator at its rated capacity which

encompasses all load demands for the Loss

of Coolant Accident concurrent with a Loss

of Offsite Power (LOCA/LOOP) scenario.

Because of reduced temperature and

pressure, LOCA/LOOP is a less significant

and probable event in MODES 5 and 6. The

bounding scenario is considered to be a Loss

of Offsite Power (LOOP) while the plant is

shutdown (in MODES 5 and 6). The new

diesel fuel oil inventory required during

plant shutdown conditions is based on

LOOP. Because this change only affects

diesel fuel inventory, there is no impact on

the probability of an accident. The

consequences of LOOP event are unchanged

since sufficient fuel remains available to

allow the diesel generators to support

mitigation of the event. Because seven days

of fuel are required, there is no change in the

consequences of any event which requires

the diesel generators. Therefore, there is no

significant increase in the probability or

consequences of an accident previously

evaluated as a result of this proposed change.

ADDITION OF REMEDIAL ACTION TO

RESTORE THE STORED VOLUME OF

DIESEL FUEL

The second proposed change applies to all

MODES of operation. This change allows the

diesel generator to remain OPERABLE if the

fuel oil inventory falls below the minimum

required in the storage system (i.e., fuel

volume for 7-day operation of the diesel

generator) but remains above a fuel volume

for 6 days operation of the diesel generator.

The minimum required fuel oil volume must

be restored within 48 hours of falling below

the limit. This relaxation by 48 hours allows

sufficient time to replenish the required fuel

oil volume and complete any required

analysis prior to fuel oil addition to the

storage tank. Because this change only affects

diesel generator fuel inventory, there is no

impact on the probability of an accident.

Since the fuel oil replenishment can be

obtained in less than six days after an event,

there is no significant increase in the

probability of a loss of all AC power (i.e.,

Station Blackout). Because the remaining fuel

oil volume is larger than 6-day fuel supply

and actions are initiated to obtain

replenishment within this brief period, the

proposed change does not involve a

significant increase in the consequences of an

accident previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

REDUCTION IN MINIMUM DIESEL FUEL

STORED VOLUME WHILE SHUTDOWN

The first proposed change reduces the

diesel fuel oil inventory required for plant

shutdown conditions. As described above,

LOOP is the limiting condition for diesel fuel

oil inventory requirements for a plant in the

shutdown condition. As the proposed fuel

inventory is adequate for a shutdown LOOP

and no hardware changes or system

operation changes are involved, no new

failure modes are introduced and hence, no

new or different accidents from any

previously evaluated are created.

ADDITION OF REMEDIAL ACTION TO

RESTORE THE STORED VOLUME OF

DIESEL FUEL

The second proposed change only affects

diesel generator fuel inventory as well. There

are no hardware changes and no changes in

system operations involved; therefore, no

new or different accidents from any accident

previously evaluated are created.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

The intent of the Technical Specification is

to conservatively assure sufficient fuel to

assure diesel generator operation to support

mitigation of postulated events. This intent is

accomplished by conservatively assuring a

seven day supply of fuel. Seven days fuel

supply is considered sufficient to support the

initial mitigation activities, identify the need

for additional fuel, arrange for delivery, test

and then add fuel to the storage tanks, if

needed. The current diesel fuel oil inventory

for operating conditions (MODES 1, 2, 3 and

4), is sufficient to conservatively support

seven days of diesel generator operation for

a LOCA with LOOP condition.

REDUCTION IN MINIMUM DIESEL FUEL

STORED VOLUME WHILE SHUTDOWN
32374 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

The proposed diesel fuel oil inventory for

shutdown conditions (MODES 5 and 6), is

adequate to conservatively support seven

days of diesel generator operation for LOOP

conditions. The proposed reduction in

inventory between operating and shutdown

conditions continues to support the different

transient conditions which are applicable to

the different modes of operation. Even

though the minimum storage requirement

during shutdown is being reduced, the basis

of this specification continues to be

conservatively satisfied and therefore this

license amendment request does not involve

a significant reduction in a margin of safety.

ADDITION OF REMEDIAL ACTION TO

RESTORE THE STORED VOLUME OF

DIESEL FUEL

The second proposed change which is

applicable to all MODES of operation, allows

48 hours to restore diesel generator fuel oil

inventory to the seven-day level as long as

the inventory does not fall below the six-day

level. The probability of a LOOP during this

period is low. The 6-day fuel oil supply is

calculated with adequate margin similar to

the calculation of 7-day fuel oil inventory. In

spite of the potential that there may be

slightly less fuel available inlenishment

within this brief period. Based on this and

the low probability of an event during this

brief period, it is considered that this change

request does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar,

Esq., Newman and Holtzinger, 1615 L

Street, N.W., Suite 1000, Washington,

D.C. 20036

NRC Project Director: William D.

Beckner

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Arizona Public Service Company, et al.,

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

and STN 50-530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

May 4, 1994

Brief description of amendments: The

amendments revise Limiting Condition

for Operation (LCO) 3.4.8.3 and

Surveillance Requirement 4.4.8.3.1,

‘‘Overpressure Protection Systems.’’

Specifically, the LCO and surveillance

requirements are revised to clarify that

both shutdown cooling system (SCS)

suction relief valves shall be OPERABLE

and aligned to provide overpressure

protection not only during reactor

coolant system (RCS) cooldown and

heatup evolutions, but also during any

steady-state temperature periods in the

course of RCS cooldown or heatup

evolutions.

Date of issuance: June 2, 1995

Effective date: June 2, 1995

Amendment Nos.: Unit 1 -

Amendment No. 93; Unit 2 -

Amendment No. 80; Unit 3 -

Amendment No. 63

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR 42333)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 2, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power

Station,Plymouth County,

Massachusetts

Date of application for amendment:

November 22, 1994

Brief description of amendment: This

amendment revises the suppression

chamber water level operating range,

increasing it 2 inches, and revises the

water level recorder range in response to

a commitment from an inspection.

Date of issuance: June 1, 1995

Effective date: June 1, 1995

Amendment No.: 163

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 18, 1995 (60 FR 3672)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Carolina Power & Light Company,

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

SteamElectric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

February 24, 1995

Brief description of amendment: The

proposed change would remove Section

4.3 from the Technical Specifications

(TS) because the primary system testing

following opening is already performed

in accordance with the American

Society of Mechanical Engineers Boiler

and Pressure Vessel Code, as

implemented in the licensee’s inservice

inspection program as required by TS

4.0.1.

Date of issuance: May 30,

1995Effective date: May 30, 1995

Amendment No.: 165

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16183)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 30, 1995.No

significant hazards consideration

comments received: No
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32375

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois; Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

October 15, 1992, as supplemented

March 9, 1993.

Brief description of amendments: The

amendments would modify the existing

Dresden and Quad Cities Technical

Specifications (TS) to format them in

the style of the Boiling Water Reactor 4

(BWR) Standard Technical

Specifications (STS). The amendments

deal specifically with Section 3/4.4,

‘‘Standby Liquid Control System

(SLCS).’’

Date of issuance: June 8, 1995

Effective date: For Dresden,

immediately, to be implemented no

later than December 31, 1995; for Quad

Cities, immediately, to be implemented

no later than June 30, 1996.

Amendment Nos.: 133, 127, 154, and

150

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 7, 1993 (58 FR 36429) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 8, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Commonwealth Edison Company,

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

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of application for amendments:

April 10, 1995

Brief description of amendments: The

amendments would change the

Technical Specifications by (1) revising

the low pressure value at which the

High Pressure Coolant Injection (HPCI)

and Reactor Core Isolation Cooling

(RCIC) systems can be tested to 150 psig,

and (2) testing these systems against a

system head corresponding to reactor

vessel pressure when steam is supplied

to the turbines at 920 psig to 1005 psig

for high pressure testing and 150 psig to

325 psig for low pressure testing.

Date of issuance: May 30, 1995

Effective date: Immediately and shall

be implemented within 60 days.

Amendment Nos.: 153 and 149

Facility Operating License Nos. DPR29 and DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 28, 1995 (60 FR 21009)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 30, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Commonwealth Edison Company,

Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

November 21, 1994.

Brief description of amendments: The

amendments add footnotes in Limiting

Condition for Operation 3.15.2.A of the

Technical Specifications (TS) to allow a

one-time extension of the allowed

outage time (AOT) for an inoperable

reserve offsite power source from 72

hours to 14 days. To provide additional

assurance that redundant sources of

power to the operating unit are operable

during the AOT outage, the amendment

also adds footnotes in Surveillance

Requirement 4.15.2.A of the TS to

modify the emergency diesel generator

and the normal offsite power source

testing requirements.

Date of issuance: May 31, 1995

Effective date: May 31, 1995

Amendment Nos.: 163 and 151

Facility Operating License Nos. DPR39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 500).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 31, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, MichiganDate of application

for amendment: October 20, 1992

Brief description of amendment: This

amendment revises Technical

Specification 5.3.1a to account for

changes being made to the Palisades

Final Safety Analysis Report (FSAR)

Section 4.2 following replacement of the

steam generators.

Date of issuance: May 22, 1995

Effective date: May 22, 1995

Amendment No.: 166

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18624)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 22, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

January 13, 1995, as supplemented

April 12 and 27, 1995

Brief description of amendment: This

amendment revises the Technical

Specifications to allow installed

primary and secondary safety valve

settings to be within a 3% tolerance of

their nominal settings, but would

require returning the valve settings to

within 1% of the nominal settings if the

valves are removed from the piping for

maintenance or testing.

Date of issuance: June 8, 1995

Effective date: June 8, 1995

Amendment No.: 167

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11130)

The April 12 and 27, 1995, letters

provided clarifying information in

response to the staff’s request for

additional information of April 11,

1995, and a telephone request for

information on the Palisades loss of load

analysis contained in the January 13,

1995, submittal. This information was

within the scope of the original

application and did not change the

staff’s initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 8, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.
32376 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan Date of application for

amendment: September 13, 1993

Brief description of amendment: The

amendment revises Technical

Specification (TS) 6.5.2.8 to relocate

audit frequencies from the TS to the

Quality Assurance Program located in

Chapter 17.2 of the Updated Final

Safety Analysis Report. A related

change to extend the frequency of the

use of an independent fire contractor to

every third fire protection audit was

denied.

Date of issuance: May 23, 1995

Effective date: May 23, 1995, with full

implementation within 45 days.

Amendment No.: 104

Facility Operating License No. NPF43. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18625)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 23, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request:

September 16, 1993

Brief description of amendment: The

amendment revised the Technical

Specifications by removing the incore

detection system requirements. These

requirements are to be relocated in the

Updated Final Safety Analysis Report.

Date of issuance: May 30, 1995

Effective date: May 30, 1995, to be

implemented within 60 days.

Amendment No.: 107

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 27, 1993 (58 FR

57851) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

May 30, 1995. No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122.

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of application for amendments:

January 20, 1995

Brief description of amendments:

These amendments will relocate the

operability requirements for Incore

Detectors in Technical Specification 3/

4.3.3.2 to the Updated Final Safety

Analysis Report, and revise Linear Heat

Rate Surveillance 4.2.1.4, and Special

Test Exceptions Surveillances 4.10.2.2,

4.10.4.2 (Unit 2 only), and 4.10.5.2,

accordingly.

Date of issuance: June 6, 1995

Effective date: June 6, 1995

Amendment Nos.: 136 and 75

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

the Technical Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11132)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 6, 1995No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-424 and 50-

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

December 29, 1994, as supplemented by

letter dated May 2, 1995.

Brief description of amendments: The

amendments revise TS 3/4.3,

Instrumentation and its associated

Bases, and TS 3/4.8, Electrical Power

Systems to specify the appropriate

actions to take in the event that an

automatic load sequencer must be taken

out of service or becomes inoperable.

Date of issuance: May 31, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 86 and 64

Facility Operating License Nos. NPF68 and NPF-81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6301).

The May 2, 1995, letter provided minor

editorial changes that did not change

the scope of the December 29, 1994,

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 31, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830.

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: January

13, 1993, as supplemented by letter

dated October 18, 1993

Brief description of amendment: The

amendment revised the River Bend

Station, Unit 1 operating license to

reflect a change in ownership of Gulf

States Utilities (GSU). GSU, which

ownes a 70 percent undivided interest

in the River Bend Station, is a whollyowned subsidiary company of Entergy

Corporation. This amendment was

originally issued on December 16, 1993,

as License Amendment No. 69.

Date of issuance: June 8, 1995.

Effective date: June 8, 1995.

Amendment No.: 78

Facility Operating License No. NPF47. The amendment revised the

operating license.

Date of initial notice in Federal

Register: July 7, 1993 (58 FR 36436) The

October 18, 1993, supplemental letter

provided clarifying information and did

not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 8, 1995.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: January

13, 1993, as supplemented by letter

dated June 29, 1993

Brief description of amendment: The

amendment revised the River Bend

Station, Unit 1 operating license to

include as a licensee, Entergy

Operations, Inc. (EOI), and to authorize

EOI to use and operate River Bend and

to possess and use related licensed

nuclear materials. This amendment was

originally issued on December 16, 1993

as License Amendment No. 70.

Date of issuance: June 8, 1995

Effective date: June 8, 1995

Amendment No.: 79
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32377

Facility Operating License No. NPF47. The amendment revised the

operating license.

Date of initial notice in Federal

Register: July 7, 1993 (58 FR 36436) The

June 29, 1993, supplemental letter

provided clarifying information and did

not change the initial no significant

hazards consideration

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

June 8, 1995.No significant hazards

consideration comments received. Yes.

Comments and a request for hearing

were received from Cajun Electric

Power Cooperative of Baton Rouge,

Louisiana.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, Louisiana 70803.

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

Duane Arnold Energy, Center, Linn

County, Iowa

Date of application for amendment:

March 1, 1995

Brief description of amendment: The

amendment revised the surveillance

criteria for certain pumps and valves in

the Low Pressure Coolant Injection

(LPCI) subsystem; the Core Spray

subsystems; and the Residual Heat

Removal (RHR) Service Water, High

Pressure Coolant Injection (HPCI),

Emergency Service Water (ESW), and

River Water Supply systems. The

surveillance criteria changed from every

three months to the testing frequency

specified in the Inservice Testing

program.

Date of issuance: May 18, 1995

Effective date: May 18, 1995

Amendment No.: 210

Facility Operating License No. DPR49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18626)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 18, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401.

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

Duane Arnold Energy, Center, Linn

County, Iowa

Date of application for amendment:

March 10, 1995

Brief description of amendment: The

amendment deletes Technical

Specification Sections 3.7/4.7.H.3 to

eliminate redundant Limiting

Conditions of Operation and

Surveillance Requirements for the

containment hydrogen and oxygen

analyzers.

Date of issuance: May 31, 1995

Effective date: May 31, 1995

Amendment No.: 211

Facility Operating License No. DPR49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20518)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 31, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401.

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

September 28, 1990

Brief description of amendment: The

amendment revised the Technical

Specifications to establish periodic

operability testing of the reactor vessel

overfill protection system. The changes

were requested to satisfy a commitment

in the licensee’s response to Generic

Letter 89-19, ‘‘Request for Action

Related to Resolution of Unresolved

Safety Issue (USI) A-47.’’

Date of issuance: June 8, 1995

Effective date: June 8, 1995

Amendment No.: 169

Facility Operating License No. DPR46. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 31, 1990 (55 FR

45885) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

June 8, 1995. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, NE 68305.

Northeast Nuclear Energy Company,

Docket No. 50-245, MillstoneNuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

March 31, 1995

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to increase the asfound setpoint tolerance of the safety/

relief valves (SRVs) from plus or minus

1% to plus or minus 3%. In addition,

the amendment (1) allows the as-found

condition of one SRV to be inoperable,

(2) clarifies the 1325 psig safety limit

wording, (3) increases the number of

SRVs to be tested during each refueling

outage, (4) makes editorial changes to

reflect the TS changes, and (5) revises

the bases for the applicable sections.

Date of issuance: May 31, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 82

Facility Operating License No. DPR21. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20520)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 31, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50-423,

MillstoneNuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of application for amendment:

January 23, 1995.

Brief description of amendment: The

amendment revises the Technical

Specifications to modify the

containment spray system by replacing

the present sodium hydroxide spray

additive with the trisodium phosphate

dodecahydrate pH control agent.

Date of issuance: May 26, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 115

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11136).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 26, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

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 application for amendments:

September 20, 1994, as supplemented

by letter dated April 14, 1995.
32378 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

Brief description of amendments: The

proposed amendments revise

surveillance requirements (SRs) as

recommended by NRC Generic Letter

(GL) 93-05, ‘‘Line-Item Technical

Specification Improvements to Reduce

Surveillance Requirements for Testing

During Power Operation’’ of the

combined Technical Specifications (TS)

for the Diablo Canyon Nuclear Power

Plant Unit Nos. 1 and 2. The specific TS

changes are as follows:

(1) TS SR 4.1.3.1.2 is revised to

change the frequency for testing the

movability of the control rods from at

least once per 31 days to at least once

per 92 days.

(2) TS 3/4.3.2, Table 4.3-2,

‘‘Engineered Safety Features Actuation

System Instrumentation Surveillance

Requirements,’’ Functional Unit 3.c.4),

and TS 3/4.3.3.1, Table 4.3-3,

‘‘Radiation Monitoring Instrumentation

for Plant Operations SRs,’’ is revised to

change the monthly channel functional

test to quarterly.

(3) TS 3/4.5.1 is changed as follows:

(a) TS SR 4.5.1.1a.1) is revised to more

clearly state that the accumulator water

volume and pressure must be verified to

be within their limits. (b) TS SR

4.5.1.1b. is revised to specify that the

boron concentration surveillance is not

required to be performed if the

accumulator makeup source was the

refueling water storage tank (RWST). (c)

TS SR 4.5.1.2 is relocated to plant

procedures.

(4) TS SR 4.5.2c.2) is revised to clarify

that a separate containment entry to

verify the absence of loose debris is not

required after each containment entry.

(5) TS SR 4.6.2.1d. is revised to

change the frequency for a containment

spray header flow test from at least once

per 5 years to at least once per 10 years.

(6) TS SR 4.6.4.2a. is revised to

change the verification of the minimum

hydrogen recombiner sheath

temperature from at least once per 6

months to at least once each refueling

interval.

(7) TS SR 4.7.1.2.1 is revised to

change the surveillance frequency for

testing each auxiliary feedwater (AFW)

pump from at least once per 31 days to

at least once per 92 days on a staggered

test basis.

(8) TS SR 4.10.1.2 is revised to

lengthen the allowed period of time for

a rod drop test from 24 hours to 7 days

prior to reducing shutdown margin to

less than the limits of TS 3.1.1.1.

(9) TS SR 4.11.2.6 is revised to change

the surveillance frequency from 24

hours to 7 days when radioactive

material is being added to the gas decay

tanks and to add a requirement to

monitor radioactive material

concentrations in the gas decay tanks at

least once per 24 hours when system

degassing operations are in progress.

Date of issuance: May 26, 1995

Effective date: May 26, 1995, to be

implemented within 60 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 102; Unit 2

Amendment No. 101

Facility Operating License Nos. DPR80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53843) The April 14, 1995, letter

provided clarifying information and did

not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 26, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

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 application for amendments:

December 30, 1994 (LAR 94-12)

Brief description of amendments:

These amendments clarify the technical

specifications (TS) issued in license

amendments 84/83 associated with the

Eagle 21 reactor protection system

modification, delete TS references to

RM-14A and RM-14B, remove cyclespecific TS requirements, and

incorporate editorial corrections.

Date of issuance: June 2, 1995

Effective date: June 2, 1995, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 103; Unit 2 -

Amendment No. 102

Facility Operating License Nos. DPR80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14026)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 2, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

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 application for amendments:

February 6, 1995, as supplemented by

letters dated March 23, 1995, and May

22, 1995.

Brief description of amendments: The

amendments would allow the storage of

fuel with enrichments up to and

including 5.0 weight percent U-235,

would clarify that substitution of fuel

rods with filler rods is acceptable for

fuel designs that have been analyzed

with applicable NRC-approved codes

and methods, and would allow the use

of ZIRLO fuel cladding in the future in

addition to Zircaloy-4.

Date of issuance: June 7, 1995

Effective date: June 7, 1995, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 104; Unit 2 -

Amendment No. 103

Facility Operating License Nos. DPR80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11138)

The licensee’s supplemental letters

provided additional clarifying

information. The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

June 7, 1995. No significant hazards

consideration comments received: Yes.

Comments were submitted by Jill

ZamEk on behalf of the San Luis Obispo

Mothers for Peace by letter dated March

30, 1995.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Pacific Gas and Electric Company,

Docket No. 50-133, Humboldt Bay

Power Plant, Unit 3, Humboldt County,

California

Date of application for amendment:

November 23, 1994, as supplemented

April 27, 1995.

Brief description of amendment: This

amendment revised the Technical

Specifications Section VII.C., Plant

Staff, to decrease the minimum staff

requirements for the shift operating

organization from five to two persons.

Date of issuance: May 31, 1995

Effective date: This license

amendment is effective as of the date of
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32379

its issuance and must be fully

implemented no later than 30 days from

the date of issuance.

Amendment No.: 28Facility License

No. DPR-7: The amendment revised the

TS.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11139)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 31, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Humboldt County Library, 636

F Street, Eureka, California 95501.

PECO Energy Company, Public Service

Electric and Gas CompanyDelmarva

Power and Light Company, and

Atlantic City Electric Company,Docket

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

Atomic Power Station, Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments:

February 10, 1995

Brief description of amendments:

These amendments correct

administrative errors in Section 4.11.A

of the Technical Specifications (TSs).

The errors were made in the TSs by

Amendments 9 and 7 dated June 25,

1975.

Date of issuance: May 30, 1995

Effective date: May 30, 1995

Amendments Nos.: 202 and 205

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20521)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 30, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

PECO Energy Company, Public Service

Electric and Gas CompanyDelmarva

Power and Light Company, and

Atlantic City Electric Company,Docket

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

Atomic Power Station,Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments:

September 26, 1994

Brief description of amendments:

These amendments extend the

surveillance test intervals and allowable

out-of service times for the testing and

or repair of instrumentation that actuate

the Reactor Protection System, Primary

Containment Isolation, Core and

Containment Cooling systems, Control

Rod Blocks, Radiation Monitoring

systems and Alternate Rod Insertion/

Recirculation Pump Trip.

Date of issuance: June 6, 1995

Effective date: June 6, 1995

Amendments Nos.: 203 and 206

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14027)

The supplemental letters dated January

5, and March 23, 1995, provided

clarifying information and did not

change the initial proposed no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

June 6, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

April 30, 1993

Brief description of amendments:

These amendments changed the

Technical Specifications by deleting

Section 3/4.3.8 of the Turbine

Overspeed Protection System.

Date of issuance: June 1, 1995

Effective date: June 1, 1995

Amendment Nos.: 146 and 116

Facility Operating License Nos. NPF14 and NPF-22. These amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 9, 1993 (58 FR 32389)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

August 25, 1993, as supplemented by

letters dated June 27, 1994, and May 5,

1995

Brief description of amendments:

These amendments modify Technical

Specification Surveillance Requirement

4.7.1.3 to require that all spray pond

spray network piping above the frost

line be drained at an ambient

temperature below 40°F, and within 1

hour after being used only when the

ambient air temperature is below 40°F.

Date of issuance: June 1, 1995

Effective date: June 1, 1995

Amendment Nos.: 90 and 54

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 29, 1993 (58 FR

50972) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

June 1, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

August 30, 1994

Brief description of amendment: The

changes relocate Technical

Specification (TS) 3.3.7.9, Loose Parts

Detection System (LPDS), Surveillance

Requirement 4.3.7.9, and associated

Bases from the TSs to the Updated Final

Safety Analysis Report. The TS index is

also revised by removing the reference

to LPDS.

Date of issuance: May 25, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 73

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16197)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 25, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

February 9, 1995

Brief description of amendments: The

amendments revise the Administrative
32380 Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices

Controls section of the Technical

Specifications to reflect organizational

changes and resultant management title

changes.

Date of issuance: June 6, 1995

Effective date: June 6, 1995

Amendment Nos.: 168 and 150

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16200)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 6, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Southern Nuclear Operating Company,

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

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama.

Date of application for amendments:

March 6, 1995

Brief description of amendments: The

amendments relocate the seismic and

meteorological monitoring

instrumentation from the Technical

Specifications to the Final Safety

Analysis Report in accordance with the

‘‘Final Policy Statement on Technical

Specifications Improvements for

Nuclear Power Reactors,’’ dated July 22,

1993.

Date of issuance: May 22, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 115 and 107

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

Technical Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18628)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 22, 1995.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.

Tennessee Valley Authority, Docket No.

50-327, Sequoyah Nuclear Plant, Unit 1,

Hamilton County, Tennessee

Date of application for amendment:

April 6, 1995 (TS 95-09)

Brief description of amendment: The

amendment modifies Operating License

Condition 2.C.(25) to provide a limited

extension of the ice condenser

surveillance test interval on Unit 1 to

coincide with the Cycle 7 refueling

outage.

Date of issuance: May 30, 1995

Effective date: May 30, 1995

Amendment No.: 200

Facility Operating License Nos. DPR77: Amendment revises the technical

specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20526)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 30, 1995.No

significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

April 6, 1995

Brief description of amendments: The

amendments revise the surveillance

requirement for the power range

neutron flux channel calibration

frequency from monthly to every 31

effective full power days and delays first

performance of the surveillance after

reaching 15 percent power for 96 hrs.

Date of issuance: May 30, 1995

Effective date: May 30, 1995

Amendment Nos.: 199 and 190

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20530)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 30, 1995.No

significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

April 6, 1995

Brief description of amendments: The

amendments revise the definition of

core alteration, quadrant power tilt

ratio, and modifies the operational

mode parameters table in the Unit 1

technical specifications.

Date of issuance: June 1, 1990

Effective date: June 1, 1990

Amendment Nos.: 201 and 191

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20531)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 1, 1995.No

significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: February

14, 1994 (TXX-94046 LAR 94-006)

Brief description of amendments: The

proposed changes would revise the

Technical Specifications (TSs) for

Comanche Peak Steam Electric Station,

Units 1 and 2 in the following three

areas: 1) a change to the allowable value

for the Unit 2 pressurizer pressure-low

and Unit 2 overtemperature N-16 (OTN16) reactor trip setpoints; 2) an

administrative change to delete an

option which allowed continued

operation for a period of time when a

reactor trip system (RTS) or engineered

safety features actuation system

(ESFAS) instrumentation or interlocks

trip setpoint is found less conservative

than the allowable value; and 3) an

administrative change to combine the

Unit 1 and Unit 2 line items for RTS or

ESFAS trip setpoint and allowable

values which are the same.

Date of issuance: May 31, 1995

Effective date: May 31, 1995, to be

implemented within 30 days.

Amendment Nos.: Unit 1 -

Amendment No. 41; Unit 2 -

Amendment No. 27

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 22, 1994 (59 FR 32238)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 31, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

24, 1995

Brief description of amendment: The

amendment relaxes the requirement to
Federal Register / Vol. 60, No. 119 / Wednesday, June 21, 1995 / Notices 32381

sample the accumulator after refilling

from the RWST.

Date of issuance: May 30, 1995

Effective date: May 30, 1995, to be

implemented within 30 days of

issuance.

Amendment No.: Amendment No. 87

Facility Operating License No. NPF42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18632)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 30, 1995.No

significant hazards consideration

comments received: No. Local Public

Document Room locations: Emporia

State University, William Allen White

Library, 1200 Commercial Street,

Emporia, Kansas 66801 and Washburn

University School of Law Library,

Topeka, Kansas 66621.

Dated at Rockville, Maryland, this 14th day

of June, 1995.

For the Nuclear Regulatory Commission

John N. Hannon,

Acting Deputy Director, Division of Reactor

Projects - III/IV, Office of Nuclear Reactor

Regulation

[Doc. 95–15057 Filed 6–20–95; 8:45]

BILLING CODE 7590–01–F

[Docket No. 50–255]

Consumers Power Company

(Palisades Plant); Exemption

I

Consumers Power Company (CPCo,

the licensee) is the holder of Facility

Operating License No. DPR–20 which

authorizes operation of the Palisades

Plant, a pressurized water reactor (PWR)

located in Van Buren County, Michigan.

The license provides, among other

things, that the facility is subject to all

rules, regulations, and orders of the

Nuclear Regulatory Commission (the

Commission) now or hereafter in effect.

II

Pursuant to 10 CFR 50.12(a), the NRC

may grant exemptions from the

requirements of the regulations (1)

which are authorized by law, will not

present an undue risk to the public

health and safety, and are consistent

with the common defense and security;

and (2) where special circumstances are

present.

Section III.D.1.(a) of Appendix J to 10

CFR part 50 requires the performance of

three Type A containment integrated

leakage rate tests (ILRTs), at

approximately equal intervals during

each 10-year service period of the

primary containment. The third test of

each set shall be conducted when the

plant is shut down for the 10-year

inservice inspection of the primary

containment.

III

By letter dated March 17, 1995, as

supplemented April 26, 1995, CPCo

requested temporary relief from the

requirement to perform a set of three

Type A tests at approximately equal

intervals during each 10-year service

period of the primary containment. The

requested exemption would permit a

one-time interval extension of the third

Type A test by approximately 21

months (from the 1995 refueling outage,

currently scheduled to begin in May

1995, to the 1997 refueling outage) and

would permit the third Type A test of

the second 10-year inservice inspection

period to not correspond with the end

of the current American Society of

Mechanical Engineers Boiler and

Pressure Vessel Code (ASME Code)

inservice inspection interval.

The licensee’s request cites the

special circumstances of 10 CFR 50.12,

paragraphs (a)(2) (ii) and (iii), as the

basis for the exemption, and states that

the exemption would eliminate a cost of

$1 million for the Type A test which is

not necessary to achieve the underlying

purpose of the rule. 10 CFR part 50

Appendix J, states that the purpose of

the Type A, B, and C tests is to assure

that leakage through the primary

containment shall not exceed the

allowable leakage rate values as

specified in the technical specifications

or associated bases. CPCo points out

that the existing Type B and C testing

programs are not being modified by this

request and will continue to effectively

detect containment leakage caused by

the degradation of active containment

isolation components as well as

containment penetrations. It has been

the experience at the Palisades Plant

that, with the exception of the 1978 test

results, during the six Type A tests

conducted from 1974 to date, any

significant containment leakage paths

are detected by the Type B and C

testing. The Type A test results have

only been confirmatory of the results of

the Type B and C test results. The

testing history, structural capability of

the containment, and the risk

assessment establish that there is

significant assurance that the extended

interval between Type A tests will not

adversely impact the leak-tight integrity

of the containment and that

performance of the Type A test is not

necessary to meet the underlying

purpose of Appendix J. The licensee

also references the proposed revision to

Appendix J which would reduce the

frequency of Type A tests.

IV

Section III.D.1.(a) of Appendix J to 10

CFR part 50 states that a set of three

Type A leakage rate tests shall be

performed at approximately equal

intervals during each 10-year service

period.

The licensee proposes an exemption

to this section which would provide a

one-time interval extension for the Type

A test by approximately 21 months. The

Commission has determined, for the

reasons discussed below, that pursuant

to 10 CFR 50.12(a)(1) this exemption is

authorized by law, will not present an

undue risk to the public health and

safety, and is consistent with the

common defense and security. The

Commission further determines that

special circumstances, as provided in 10

CFR 50.12(a)(2) (ii) and (iii), are present

justifying the exemption; namely, that

application of the regulation in the

particular circumstances is not

necessary to achieve the underlying

purpose of the rule and would impose

excessive cost.

The underlying purpose of the

requirement to perform Type A

containment leak rate tests at intervals

during the 10-year service period is to

ensure that any potential leakage

pathways through the containment

boundary are identified within a time

span that prevents significant

degradation from continuing or

becoming unknown. The NRC staff has

reviewed the basis and supporting

information provided by the licensee in

the exemption request. The NRC staff

has noted that the licensee has a good

record of ensuring a leak-tight

containment following the submittal of

its Corrective Action Plan on June 30,

1986. The Corrective Action Plan was

submitting following three consecutive

Type A test failures, of which one was

the 1978 test failure. However, the

licensee has noted that the containment

penetration local leak rate tests (LLRT,

Type B and C tests) accounted for the

majority of the before maintenance

adjustment to the as-found ILRT (Type

A) results during the as-found test

failures. The penetration associated

with the 1978 test failure was

significantly modified in the mid-1980’s

to improve the LLRT test configuration

to properly monitor the entire

penetration boundary. In addition, the

licensee aggressively replaced or

repaired the valves and penetrations

that accounted for the as-found test

failures, with no repeat occurrences.

The NRC staff reviewed the LLRT

Corrective Action Plan and granted an