Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 121

will use their best efforts to keep each

other informed of proposed changes in

their respective rules and regulations

and licensing, inspection and

enforcement policies and criteria, and to

obtain the comments and assistance of

the other party thereon.

Article VII

The Commission and the

Commonwealth agree that it is desirable

to provide reciprocal recognition of

licenses for the materials listed in

Article I licensed by the other party or

by any other Agreement State.

Accordingly, the Commission and the

State agree to use their best efforts to

develop appropriate rules, regulations,

and procedures by which such

reciprocity will be accorded.

Article VIII

The Commission, upon its own

initiative after reasonable notice and

opportunity for hearing to the

Commonwealth, or upon request of the

Governor of the Commonwealth, may

terminate or suspend all or part of this

Agreement and reassert the licensing

and regulatory authority vested in it

under the Act if the Commission finds

that (1) such termination or suspension

is required to protect public health and

safety, or (2) the Commonwealth has not

complied with one or more of the

requirements of Section 274 of the Act.

The Commission may also, pursuant to

Section 274j of the Act, temporarily

suspend all or part of this Agreement if,

in the judgement of the Commission, an

emergency situation exists requiring

immediate action to protect public

health and safety and the

Commonwealth has failed to take

necessary steps. The Commission shall

periodically review this Agreement and

actions taken by the Commonwealth

under this Agreement to ensure

compliance with Section 274 of the Act.

Article IX

This Agreement shall become

effective on [April 24, 1996,] >(date to

be determined)< and shall remain in

effect unless and until such time as it is

terminated pursuant to Article VIII.

Done at [Boston, Massachusetts]

>(location to be determined)<, in

triplicate, this [24]th Day of [April,

1996] >(date to be determined)<.

For the United States Nuclear Regulatory

Commission.

lllllllllllllllllllll

Shirley Ann Jackson,

Chairman.

For the Commonwealth of Massachusetts.

lllllllllllllllllllll

William F. Weld,

Governor.

[FR Doc. 96–33252 Filed 12–31–96; 8:45 am]

BILLING CODE 7590–01–P

Sunshine Act Meeting

DATE: Weeks of December 30, 1996 and

January 6, 13, and 20, 1997.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED:

Week of December 30

There are no meetings scheduled for

the Week of December 30.

Week of January 6–Tentative

Tuesday, January 7

9:30 a.m. Briefing on Investigative

Matters (Closed—Ex. 5 & 7)

2:00 p.m. Discussion of Procedures for

NRC Strategic Assessment

(Closed—Ex. 2)

Thursday, January 9

10:00 a.m. Briefing by Maine Yankee,

NRR, and Region I (PUBLIC

MEETING) (Contact: Daniel

Dorman, 301–415–1429)

12:00 am. Affirmation Session (PUBLIC

MEETING) (if needed)

Week of January 13–Tentative

Monday, January 13

10:00 a.m. Briefing on NRC Strategic

Assessment (PUBLIC MEETING)

(Contact: John Craig, 301–415–

3812)

11:30 a.m. Affirmation Session

(PUBLIC MEETING) (if needed)

Week of January 20–Tentative

Tuesday, January 21

3:30 p.m. Briefing on Investigative

Matters (Closed—Ex, 5 & 7)

Wednesday, January 22

10:00 a.m. Briefing on Codes and

Standards (PUBLIC MEETING)

(Contact: Gil Millman, 301–415–

5843)

11:30 a.m. Affirmation Session

(PUBLIC MEETING) (if needed)

*THE SCHEDULE FOR COMMISSION

MEETINGS IS SUBJECT TO CHANGE

ON SHORT NOTICE. TO VERIFY THE

STATUS OF MEETINGS CALL

(RECORDING)—(301) 415–1292.

CONTACT PERSON FOR MORE

INFORMATION: Bill Hill (301) 415–

1661.

* * * * *

The NRC Commission Meeting

Schedule can be found on the Internet

at: http://www.nrc.gov/SECY/smj/

schedule.htm

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661).

In addition, distribution of this

meeting notice over the internet system

is available. If you are interested in

receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

* * * * *

Dated: December 27, 1996.

[FR Doc. 97–00063 Filed 12–30–96; 12:45

pm]

BILLING CODE 7590–01–M

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 December 9,

1996, through December 19, 1996. The

last biweekly notice was published on

December 18, 1996.

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
122 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By February 3, 1997, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 123

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

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 amendments request:

November 6, 1996

Description of amendments request:

The proposed amendment would

modify the technical specifications (TS)

to require manual blocking of one train

of fast bus transfer (FBT) within the first

hour of degraded switchyard voltage

should the switchyard voltage fall below

the level necessary for the electrical

distribution system (EDS) degraded

voltage protection to maintain

compliance with General Design Criteria

(GDC) 17. The proposed amendment

would further require the starting,

paralleling with the grid, loading, and

then separating from the grid the other

train’s emergency diesel generator (EDG)

within the first hour, rather than the

current TS which allows two hours after

onset of a degraded switchyard voltage

condition to start the EDG.

Alternatively, fast bus transfer can be

blocked in both trains within the first

hour. The proposed amendment

includes changes to the applicable notes

to reflect that these changes are no

longer temporary, but will remain as

part of the long-term solution to this

issue.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee provided its analysis of the

issue of no significant hazards

consideration. The NRC staff’s analysis

is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change reduces the

amount of time the second train of

electrical equipment is allowed to

remain in nonconformance with GDC 17

in the TS action statement. This change

only affects equipment used to mitigate

an event, and does not affect equipment

assumed to initiate any event. Thus the

probability of an accident previously

evaluated is not affected.

The proposed change brings the second

EDS train into compliance with GDC 17 at

least one hour sooner than the current TS.

Once in conformance with GDC 17, the

consequences of accidents previously

evaluated conform to the current analysis.

Thus the proposed change does not increase

the consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change only affects

equipment designed to mitigate the

effects of an accident. The proposed

change ensures that safety equipment is

configured as assumed in the current

accident analysis. The proposed change

does not affect the conditions of

structures, systems, or components

assumed in the safety analysis beyond

the existing design basis as maintained

by the current TS. The proposed change

does not, therefore, create the possibility

of a new or different kind of accident

from any accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety affected by the

proposed change is based on calculated

offsite dose consequences for postulated

transients and accidents for which the EDS

provides power for equipment required to

mitigate. The proposed change reduces the

time that one train of the EDS is allowed to

remain in nonconformance with GDC 17,

thus increasing the availability of the EDS

prior to the onset of a postulated accident

compared to the current TS. Thus the

proposed change does not increase

thecalculated offsite dose, and therefore the

proposed change does not involve a

significant reduction in a margin of safety.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendments request involve no

significant hazards consideration.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072-3999

NRC Project Director: William H.

Bateman

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

November 26, 1996

Description of amendments request:

The proposed amendment will adopt

Option B of 10 CFR Part 50, Appendix

J, to require Type B and Type C

containment leakage rate testing to be

performed on a performance-based

testing schedule. Containment leakage

rate testing is currently performed in

accordance with 10 CFR Part 50,

Appendix J, Option A, ‘‘Primary Reactor

Containment Leakage Testing for WaterCooled Power Reactors.’’ Appendix J

specifies containment leakage testing

requirements, including the types of

tests required, frequency of testing, and

reporting requirements. Containment

leakage test requirements include

performance of Integrated Leakage Rate

Tests, also known as Type A tests,

which measure overall leakage rate of

the containment; and Local Leakage

Rate Tests, also known as Types B and

C tests, which measure the leakage

through containment penetrations and

valves. The Nuclear Regulatory

Commission (NRC) has amended the

regulations to provide an alternate

performance-based option, Option B, to

the existing Appendix J. Baltimore Gas

and Electric Company (BGE) received

approval to adopt Option B for Type A

testing only. At this time, BGE plans to

adopt Option B for Types B and C

testing, as well.

BGE is revising the Containment

Leakage Rate Testing Program for Type

A testing to implement Types B and C

testing of the containment as required

by 10 CFR 50.54(o) and 10 CFR Part 50,

Appendix J, Option B. The revised

program will be developed in

accordance with the guidelines

contained in Regulatory Guide 1.163

‘‘Performance-Based Containment LeakRate Test Program,’’ dated September

1995, including errata.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.
124 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

Containment leakage rate testing is

performed in accordance with 10 CFR Part

50, Appendix J, ‘‘Primary Reactor

Containment Leakage Testing for WaterCooled Power Reactors.’’ The Appendix J

containment leakage test requirements

include performance of Type A tests, which

measure the overall leakage rate of the

containment, and Types B and C tests, which

measure the leakage through containment

penetrations and valves. The Nuclear

Regulatory Commission has amended the

regulations to provide a performance-based

alternative, Option B, to the existing

Appendix J. Baltimore Gas and Electric

Company adopted Option B for Type A

testing during the Unit 1 refueling outage

earlier this year. At this time, BGE plans to

adopt Option B for Types B and C testing.

Implementation of Option B involves no

physical or operational changes to the plant

structures, systems, or components.

Furthermore, leakage rate does not contribute

to the initiation of any postulated accidents;

therefore, this proposed change does not

involve an increase in the probability of any

previously evaluated accidents.

Types B and C testing is necessary to

demonstrate that leakage through the

containment penetrations is within the limits

assumed in the accident analyses. The only

potential effect of the proposed change to the

Types B and C test frequency is the

possibility that containment penetration

leakage would go undetected between tests.

To provide assurance that containment

penetration leakage remains within the limits

of the Technical Specifications, BGE plans to

implement the performance-based leakage

testing program in accordance with NRC

Regulatory Guide 1.163, dated September

1995 (including errata), with no exceptions.

By adopting Option B, BGE will no longer

require an exemption from 10 CFR Part 50,

Appendix J, which was granted to

accommodate 24-month operating cycles.

The exemption increased the surveillance

interval to a maximum of 30 months, while

proportionately decreasing the combined

Types B and C leakage rate acceptance

criteria. Option B to Appendix J provides the

regulation necessary to accommodate an

extended fuel cycle, while maintaining the

original combined Types B and C leakage rate

testing limit. Therefore, BGE has requested

revocation of the exemption to 10 CFR Part

50, Appendix J, as adoption of Option B for

Types B and C testing will enable a return

to full compliance with Appendix J. As the

facility will be in full compliance with the

regulations, this change does not increase the

consequences of any previously evaluated

accidents.

Implementation of Option B does not

change the total allowable containment

leakage rate acceptance criteria, nor does it

change the total leakage assumed in the

accident analyses. Option B allows the

implementation of a performance-based

testing program to ensure that resources are

concentrated on the components most likely

to exceed administrative limits. Similarly,

the changes to relocate the procedural

details, including test frequency,

performance and data conversion

methodology, for containment leakage rate

testing from the Technical Specifications to

the Containment Leakage Rate Testing

Program will have no effect on the total

containment leakage allowed by the

Technical Specifications, or assumed in the

accident analyses. Relocating the allowable

leakage rate conversions (Standard Cubic

Centimeters per Minute) to the Technical

Specification Bases does not change the

allowable leakage rates (as a percentage of the

containment air volume) specified in the

Technical Specifications. Furthermore,

relocation of the programmatic controls for

Types B and C testing, including the

allowable leakage rates, to the Administrative

Controls section of the Technical

Specifications ensures an adequate level of

regulatory control of these criteria is retained.

Additionally, the Calvert Cliffs Individual

Plant Examination considered the effects

associated with severe accidents which could

lead to containment failure. It was concluded

that adopting a performance-based testing

interval will not significantly affect the

containment failure probabilities calculated

for the Individual Plant Examination.

Altogether, adoption of a performance-based

testing frequency, as specified in 10 CFR Part

50, Appendix J, Option B, will not

significantly decrease the confidence in the

leak-tightness of the containment, including

containment penetrations. Therefore, this

change will not result in a significant

increase in the probability of undetected

containment penetration leakage in excess of

that allowed by the Containment Leakage

Rate Testing Program, or assumed in the

accident analysis, or in the consequences of

an accident previously evaluated.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

The proposed Technical Specification

change adopts a performance-based approach

to containment penetration leakage rate

testing. This change does not add any new

equipment, modify any interfaces with any

existing equipment, or change the

equipment’s function, or the method of

operating the equipment. The proposed

change does not affect normal plant

operations or configuration, nor does it affect

leakage rate test methods. As the proposed

change would not change the design,

configuration or operation of the plant, it

could not cause containment penetration

leakage rate testing to become an accident

initiator.

Therefore, the proposed change does not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The purpose of the existing schedule for

Types B and C tests is to provide assurance,

on a regular basis, that the release of

radioactive material will be restricted to

those leak paths and leakage rates assumed

in the accident analyses. The margin of safety

associated with containment penetration

leakage rates is not reduced if containment

leakage does not exceed the maximum

allowable leakage rate defined in the

Technical Specifications. Implementation of

Option B does not change the total allowable

containment leakage rate acceptance criteria,

nor does it change the total leakage assumed

in the accident analyses. Option B only

allows the implementation of a performancebased testing program to ensure that

resources are concentrated on the

components most likely to exceed

administrative limits. Similarly, the changes

to relocate the procedural details for

containment leakage rate testing from the

Technical Specifications to either the

Containment Leakage Rate Testing Program

or the Technical Specification Bases will

have no effect on the total containment

leakage allowed by the Technical

Specifications, or assumed in the accident

analyses. Furthermore, relocation of the

programmatic controls for Types B and C

testing, including the allowable leakage rates,

to the Administrative Controls section of the

Technical Specifications ensures that the

same regulatory control of these criteria is

retained.

Elimination of the exemption to Appendix

J which reduced the amount of combined

Types B and C testing allowable leakage

redistributes that portion of the total

containment leakage which may be attributed

to local leakage rate testing, but does not

affect the maximum allowable containment

leakage rate, La. The proposed change does

not affect a safety limit, a Limiting Condition

for Operation, or the way in which the plant

is operated.

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 amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: S. Singh Bajwa,

Acting Director

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan

Date of amendment request:

December 2, 1996 (NRC-96-0134)

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3.1.4.3,

Rod Block Monitor, and Tables 3.3.6-1

and 4.3.6-1 in TS 3.3.6, Control Rod

Block Instrumentation, to expand the

range of conditions under which the rod

block monitor must be operable. These

changes are required to ensure that all

fuel limits are met for the core that has

been loaded for Cycle 6.
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 125

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes provide

requirements that are more restrictive than

the existing requirements for operation of the

facility. These changes provide assurance

that the Rod Block Monitor system is

operable when necessary to prevent or

mitigate transients that could potentially

threaten the integrity of the fuel cladding.

There will be no adverse impact on the

probability of any accident previously

evaluated since the change provides

additional assurance that fuel thermal and

mechanical design bases will be satisfied and

has no effect on any accident initiating

mechanism. The additional restrictive

conditions on plant operation also ensure

that the consequences of anticipated

operational occurrences are no more severe

than the most limiting conditions using the

current Technical Specifications. Therefore

these changes do not involve any increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed changes will not involve

any physical changes to plant systems,

structures, or components (SSC). The

changes in Rod Block Monitor operability

requirements are consistent with the current

safety analysis assumptions. These

requirements provide assurance that the Rod

Block Monitor will be operable if necessary

to terminate a rod withdrawal error so that

fuel thermal and mechanical design limits

are satisfied. The change does not cause a

physical change to the plant or introduce a

new mode of operation. Therefore, the

proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. These changes maintain current

assumptions within the safety analyses and

design basis. The changes provide assurance

that the Rod Block Monitor will be operable

if necessary to terminate a rod withdrawal

error so that fuel thermal and mechanical

design limits are satisfied. Therefore, these

changes do not involve a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226

NRC Project Director: John N. Hannon

Entergy Gulf States, Inc., 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:

November 6, 1996

Description of amendment request:

The proposed amendment would revise

the technical specifications to permit an

increase in the allowable leak rate for

the Main Steam Isolation Valves

(MSIVs) and delete the Penetration

Valve Leakage Control System (PVLCS)

and Main Steam-Positive Leakage

Control System (MS-PLCS)

requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The operation of River Bend Station, in

accordance with the proposed amendment,

will not involve a significant increase in the

probability or consequences of any accident

previously evaluated.

The proposed amendment to delete

Technical Specification 3.6.1.8 and 3.6.1.9

involves eliminating the PVLCS and MSPLCS leakage control requirements from the

Technical Specifications. As described in

Sections 9.3 and 6.7 respectively, of the

USAR [Updated Safety Analysis Report], the

PVLCS and MS-PLCS are manually initialed

about 20 minutes following a design basis

LOCA [Loss of Coolant Accident].

Since the PVLCS and MS-PLCS are

operated only after an accident has occurred,

this proposed amendment has no effect on

the probability of an accident.

Since MSIV leakage and operation of the

PVLCS and MS-PLCS are included in the

radiological analysis for the design basis

LOCA as described in Section 15.6.5 of the

USAR, the proposed amendments will not

affect the precursors of other analyzed

accidents. The PVLCS and MS-PLCS are not

initiators of any previously analyzed

accident. The proposed amendments result

in acceptable radiological consequences of

the design basis LOCA previously evaluated

in Section 15.6.5 of the USAR.

The proposed amendment to Technical

Specification 3.6.1.3 does not involve a

change to structures, components or systems

that would affect the probability of an

accident previously evaluated. A plantspecific radiological analysis has been

performed to assess the affects of the

proposed increase to the allowable MSIV leak

rate and deletion of the PVLCS and MS-PLCS

in terms of Control Room and off-site doses

following a postulated design basis LOCA.

This change required a revision to the

existing LOCA dose analysis due to the

potential leakage from the MSIVs and those

valves served by the PVLCS. Additional

changes were also included in the revised

dose analysis to account for changes in

regulatory guidance and dose methodology.

Leakage from the drywell to the atmosphere

through the PVLCS (secondary containment

bypass valves) are both assumed to begin at

time zero. The model conservatively assumes

that one inboard MSIV fails open at time zero

and the MSIVs associated with the remaining

three main steam lines are assumed to begin

leakage at 2 hours with a total leak rate of

200 scfh for all four main steam lines. The

design basis leak rate of the primary

containment (excluding main steam lines and

lines sealed by the PVLCS) is 0.26% of the

containment volume by weight per 24 hours

for the duration of the accident and is

assumed to be released entirely to the

environment initially or the secondary

containment later into the accident. The

leakage of 170,000 cc/hr (4298 sccm) at Pa

through the containment isolation valves

served by the PVLCS is considered as bypass

leakage circumventing the secondary

containment. The on-site and off-site doses

were determined using the TRANSACT

computer code which included the ICRP 30

dose conversion factors. The total off-site and

on-site LOCA doses for both the airborne and

liquid release pathways resulting from the

proposed change are bounded by the

applicable regulatory limits.

The analysis demonstrates that dose

contributions from the proposed combined

MSIV leakage rate limit of 200 scfh and from

the proposed deletion of the PVLCS and MSPLCS result in values bounded by the

applicable regulatory limits as compared to

the LOCA doses previously evaluated for the

off-site and Control Room doses as contained

in 10CFR100 and 10CFR50, Appendix A

(General Design Criteria 19), respectively.

The LOCA doses previously evaluated are

discussed in Section 15.6.5 of the USAR.

The whole body (DDE [Deep Dose

Equivalent]) doses at the Low Population

Zone (LPZ) is 2.82 Rem and the Control

Room is 0.43 Rem. These values are

acceptable since the revised doses are

bounded by the Regulatory Guidelines (2.82

versus 25 Rem at the LPZ and 0.43 versus 5

Rem at the Control Room). The associated

whole boy (DDE) dose at the exclusion area

boundary (EAB) is 4.69 Rem which also

remains bounded by the Regulatory

Guideline of 25 Rem.

The thyroid CEDE [Committed Effective

Dose Equivalent] dose at the LPZ is 62.58

Rem. This is acceptable since the revised

dose of 62.58 Rem is significantly less than

the Regulatory Guideline (300 Rem). The

EAB thyroid CEDE dose is 37.53 Rem,

whereas the Control Room thyroid CEDE

dose is 11.18 Rem. These values are also

acceptable since the revised doses are well

within the Regulatory Guidelines (37.53

versus 300 Rem at the EAB and 11.18 versus

30 Rem at the Control Room). The Control

Room beta (SDE [Shallow Dose Equivalent])

dose is 9.15 Rem which also remains

bounded by the Regulatory Guideline of 30

Rem.

In summary, the proposed changes do not

result in an increase to the radiological

consequences of a LOCA previously

evaluated in the USAR. The revised LOCA

doses are bounded by the Regulatory

Guidelines. The effectiveness of the proposed

request even for leakage rates greater than the
126 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

proposed MSIV allowable leak rate ensures

that off-site and Control Room dose limits are

not exceeded.

There is no physical change to the ADS/

SRVs [Automatic Depressurization System/

Safety Relief Valve]. The PVLCS accumulator

tanks remain the backup air supply to the

ADS/SRV accumulators. A qualified longterm backup air supply remains but is

supplied from a difference source. Therefore,

the proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change deletes the

requirements for the LCS [Leakage Control

System] isolation valves which are nonPCIVs. These valves are eliminated and will

not be performing a safety function. The LCS

lines that are connected to the PCIVs and

process piping will be welded and/or capped

closed to assure primary containment

integrity is maintained. The welding and

post-weld examination procedures will be in

accordance with the American Society of

Mechanical Engineers (ASME) Code, Section

XI requirements. These welds and/or caps

will be periodically tested as part of the

primary Containment Integrated Leak Rate

Test (CILRT) program in accordance with the

requirements of 10CFR50, Appendix J. The

proposed change does not involve an

increase in the probability of equipment

malfunction previously evaluated in the

USAR. In fact, the proposed change reduces

the probability of equipment malfunction

since, upon implementation, RBS will be

operated with fewer process line isolation

valves and associated support equipment

subjected to postulated failure. The affected

LCS MOVs [Motor Operated Valves] will be

eliminated or retained as normal system

isolation or maintenance valves having no

safety or leakage control function thus

requiring no bypassing of their thermal

overloads. This proposed change has no

effect on the consequences of an accident

previously evaluated since the LCS lines will

be welded and/or capped closed, thus

assuring that primary containment integrity,

isolation and leak test capability are not

compromised.

Therefore, as discussed above, the

proposed changes do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

(2) The operation of River Bend Station, in

accordance with the proposed amendment,

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed amendment to Technical

Specification 3.6.1.3 does not create the

possibility for a new or different kind of

accident from any accident previously

evaluated. The BWROG (Boiling Water

Reactors Owners Group) evaluated MSIV

leakage performance and concluded that

MSIV leakage rates up to 200 scfh will not

inhibit the capability and isolation

performance of the valve to isolate the

primary containment. There is no new

modification which could impact the MSIV

operability. The LOCA has been reanalyzed

at the proposed maximum combined leakage

rate of 200 scfh. Therefore, the proposed

change does not create any new or different

kind of accident from any accident

previously evaluated in the USAR.

The proposed amendment to delete

Technical Specification 3.6.1.8 and 3.6.1.9

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the removal of

the PVLCS and MS-PLCS does not affect any

of the remaining systems at RBS [River Bend

Station) and the LOCA has been reanalyzed

with LOCA doses resulting from the

proposed change remaining bounded by the

applicable regulatory limits.

The PVLCS and MS-PLCS are of low safety

significance as discussed in NUREG-1273,

Technical Findings and Regulatory Analysis

for Generic Safety Issue II.E.4.3,

‘‘Containment Integrity Check,’’ and NUREG/

CR-3539, ‘‘Impact of Containment Building

Leakage on LWR Accident Risk.’’

The proposed change to eliminate the LCS

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the removal of

the LCS does not adversely affect any of the

remaining RBS systems or change system

inter-relationships. The associated proposed

changes to delete the LCS isolation valves

does not create the possibility of a new or

different kind of accident. The affected LCS

MOVs will be eliminated or retained as

normal system isolation or maintenance

valves having no safety or leakage control

function thus requiring no bypassing of their

thermal overloads. The PVLCS and MS-PLCS

connections to the process piping will be

welded and/or capped closed to assure that

primary containment integrity, isolation and

leak testing capability are not compromised,

therefore eliminating the possibility for any

new or different kind of accident.

Therefore, as discussed above, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) The operation of River Bend Station, in

accordance with the proposed amendment,

will not involve a significant reduction in a

margin of safety.

The proposed amendment to Technical

Specification 3.6.1.3 does not involve a

significant reduction in a margin of safety.

The allowable leak rate limit specified for the

MSIVs is used to quantify a maximum

amount of bypass leakage assumed in the

LOCA radiological analysis. Results of the

analysis demonstrate calculated doses,

assuming the two single active failures of one

MSIV to close and one diesel generator to

respond are bounded by the requirements of

10CFR100 for the off-site doses and 10CFR50,

Appendix A (General Design Criteria 19) for

the Control Room doses. The calculated

whole body doses are significantly reduced at

the LPZ, the Control Room, and the EAB. The

calculated thyroid dose is significantly

reduced at the LPZ, the Control Room, and

the EAB.

The proposed amendment to delete

Technical Specification 3.6.1.8 and 3.6.1.9

for the PVLCS and MS-PLCS, does not reduce

the margin of safety. In fact, the overall

margin of safety is increased. The method is

effective to reduce dose consequences of

MSIV and the PVLCS leakage over an

expanded operating range and will, thereby,

resolve the safety concern that the PVLCS

and MS-PLCS will not function at leakage

rates higher than their design capacity. The

method is consistent with the philosophy of

protection by multiple leak-tight barriers

used in containment design for limiting

fission product release to the environment.

Therefore, the proposed method is highly

reliable and effective for MSIV leakage and

deletion of the PVLCS and MS-PLCS.

The calculation shows that MSIV leakage

rates up to 100 scfh per steam line would not

exceed the regulatory limits. Therefore, the

proposed method provides a substantial

safety margin for mitigating the radiological

consequences of MSIV leakage beyond the

proposed Technical Specification leak rate

limit of 200 scfh for all four main steam lines

(combined maximum pathway).

Minor increases in containment leakage

such as the leakage through the MSIVs, as

identified in NUREG-1273, NUREG/CR-3539,

and NUREG-1493 have been found to have

no significant impact on the risk to the

public. Therefore, the proposed change does

not result in a significant reduction in a

margin of safety.

The proposed change to delete the LCS

isolation valves does not reduce the margin

of safety. Welded and/or capped closure of

the LCS lines assure that primary

containment integrity and leak testing

capability are not compromised. The affected

LCS MOVs will be eliminated or retained as

normal system isolation or maintenance

valves having no safety or leakage control

function thus requiring no bypassing of their

thermal overloads. The PVLCS and MS-PLCS

connections to the process piping will be

welded and/or capped closed to assure that

primary containment integrity, isolation and

leak testing capability are not compromised,

therefore eliminating the possibility for a

significant reduction in the margin of safety.

Therefore, as discussed above, 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: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 127

Entergy Gulf States, Inc., 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:

November 15, 1996

Description of amendment request:

The proposed amendment would revise

the technical specifications to allow the

performance of the 24-hour emergency

diesel generator (EDG) maintenance run

while the unit is in either Mode 1 or

Mode 2. This test for the River Bend

Station (RBS) is currently prohibited in

Mode 1 and Mode 2 and allowed in

Modes 3, 4, and 5.

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

significantly increase the probability or

consequences of an accident previously

evaluated.

The RBS SAR [Safety Analysis Report]

assumes that the AC [Alternating Current]

electrical power sources are designed to

provide sufficient capacity, capability,

redundancy and reliability to ensure that the

fuel, reactor coolant system and containment

design limits are not exceeded during an

assumed design basis event. Specifically, the

SAR assumes that the onsite EDGs provide

emergency power in the event offsite power

is lost to either one or all three EDF

[Engineered Safety Features]

buses. In the event of a loss of preferred

power, the ESF electrical loads are

automatically connected to the EDGs in

sufficient time to provide for safe reactor

shutdown and to mitigate the consequences

of a design basis accident such as a LOCA

[Loss of Coolant Accident].

The proposed change to permit the 24-hour

testing of the EDGs during power operation

does not significantly increase the probability

or consequences of any previously evaluated

accident. The capability of the EDGs to

supply power in a timely manner will not be

compromised by permitting performance of

EDG testing during periods of power

operation. Design features of the EDGs and

electrical systems ensure that if a LOCA or

LOP [Loss of Offsite Power] signal, either

individually or concurrently, should occur

during testing, the EDG would be returned to

its ready-to-load condition (i.e., EDG running

at rated speed and voltage separated from the

offsite sources) or separately connected to the

ESF bus providing ESF loads. An EDG being

tested is considered to be operable and fully

capable of meeting its intended design

function. Additionally, the testing of an EDG

is not a precursor to any preciously evaluated

accidents.

If, during the test period, the EDG were to

receive a normal operation protective trip

resulting in the actuation of a generator

lockout signal, the lockout could be reset by

the operators monitoring the test. The

resulting delay does not present an

immediate challenge to the fuel cladding

integrity, reactor water level control or to

containment parameters, as demonstrated by

the bounding four-hour station blackout

coping analysis contained in RBS’s station

blackout conformance report.

Therefore, the proposed change allowing

testing of EDGs during power operation will

not significantly increase the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

As previously discussed, the proposed

change to permit the performance of EDG

testing during power operation will not affect

the operation of any system or alter any

system’s response to previously evaluated

design basis events. The EDGs will

automatically transfer from the test

configuration to the ready-to-load

configuration following receipt of a valid

signal (i.e., LOCA or LOP). In the ready-toload configuration the EDG will be running

at rated speed and voltage, separated from

the offsite source and capable of

automatically supplying power to the ESF

buses in the event that preferred power is

actually lost.

The proposed change is also the same

configuration currently used for the monthly

one-hour test. Therefore, testing during

power operation will not create the

possibility of a new or different kind of event

from any previously evaluated.

[Surveillance Requirement] SR 3.8.1.16

demonstrated that the EDG will

automatically override the test mode

following generation of a LOCA signal. In

addition, the ability of the EDGs to survive

a full load reject is verified by the

performance of SR 3.8.1.9. These existing

surveillance requirements, along with system

design features, ensure that the performance

of EDG testing during power operation will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The AC electrical power sources are

designed to provide sufficient capacity,

capability, redundancy, and reliability to

ensure the availability of necessary power to

ESF systems so that the fuel, reactor coolant

system and containment design limits are not

exceeded. Specifically, the EDGs must be

capable of automatically providing power to

ESF loads in sufficient time to provide for

safe reactor shutdown and to mitigate the

consequences of a design basis accident in

the event of a loss of preferred power.

Testing of EDGs during power operation

will not affect the availability or operation of

any offsite source of power. In addition, the

EDG being tested remains capable of meeting

it intended design functions. Therefore, the

proposed change to the Technical

Specification Surveillance Requirement

3.8.1.13 will not result in a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Entergy Gulf States, Inc., 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:

November 15, 1996

Description of amendment request:

The proposed amendment would

increase the two recirculation loop

Minimum Critical Power Ratio (MCPR)

limit from 1.07 to 1.10 and the single

recirculation loop MCPR limit from 1.08

to 1.12. This change request is the result

of a non-conservative calculation

identified by the fuel vendor.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The request does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The revised Safety Limit MCPR and the

cycle-specific thermal limits that are based

on the revised SLMCPR have been calculated

using the methods identified in the

‘‘Supplemental Reload Licensing Report For

River Bend Station Reload 6 Cycle 7’’

(Reference 1). These methods are within the

existing design and licensing basis and

cannot increase the probability or severity of

an accident. The basis of the MCPR Safety

Limit calculation is to ensure that greater that

[than] 99.9% of all fuel rods in the core avoid

transition boiling and fuel damage in the

event of a postulated accident.

The SLMCPR is used to establish the

Operating Limit Minimum Critical Power

Ratio (OLMCPR). Neither the SLMCPR nor

the OLMCPR can initiate an event,

therefore[,] a change to the SLMCPR does not

increase the probability of a accident

previously evaluated. Maintaining the

Minimum Critical Power Ratio (MCPR) at or

above the OLMCPR during normal operations

precludes fuel failure due to overheating of

the fuel clad during an anticipated

operational occurrence (AOO), thus limiting

the consequences of an AOO. The proposed

change will increase the SLMCPR, which

will require the OLMCPR to be increased,
128 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

which in turn will ensure that the

requirements of 10 CFR [Part] 100 are met for

an AOO. Therefore, there is no increase in

the consequences of an accident previously

analyzed.

The request does not create the possibility

of occurrence of a new or different kind of

accident from any accident previously

evaluated.

The MCPR Safety Limit is a Technical

Specification numerical value designed to

ensure that fuel damage from transition

boiling does not occur as a result of the

limiting postulated accident. It cannot create

the possibility of any new type of accident.

Neither the SLMCPR or the OLMCPR can

initiate an event, therefore, a change to the

SLMCPR does not create the possibility of

occurrence of a new or different kind of

accident from any accident previously

evaluated.

The request does not involve a significant

reduction in the margin of safety.

The MCPR Safety Limit is a Technical

Specification numerical value designed to

ensure that fuel damage from transition

boiling does not occur as a result of the

limiting postulated accident. This new Safety

Limit MCPR is calculated using the methods

identified in the reference. These methods

are within the existing design and licensing

basis and based on RBS specific inputs.

The margin of Safety resides between the

SLMCPR and the point at which fuel fails.

The proposed change to SLMCPR (and the

OLMCPR) will in fact restore the margin of

safety associated with GE’s SLMCPR

methodology.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam ElectricStation, Unit 3,

St. Charles Parish, Louisiana

Date of amendment request:

December 2, 1996

Description of amendment request:

The proposed Technical Specification

(TS) Change Request will permit the use

of 10CFR50 Appendix J, Option B,

Performance-Based Containment

Leakage Testing for Type A, B and C

leak rate testing. TSs 3/4.6.1.1, 3/4.6.1.2,

3/4.6.1.3, 4.6.1.6 and 4.6.1.7 are revised

and Section 6.15 is added establishing

the Containment Leakage Rate Testing

Program. The Bases are revised to reflect

this change. Minor editorial changes are

included in this request. Waterford

Steam Electric Station is planning to

have a Containment Leakage Rate

Testing Program in place prior to the

next scheduled refueling outage. This

program will be in accordance with the

guidelines contained in Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leak-Test Program,’’ dated

September 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change will not affect the

assumptions, design parameters, or results of

any accident previously evaluated. The

proposed change does not add or modify any

existing equipment. The proposed changes

will result in increased intervals between

containment leakage tests determined

through a performance based approach. The

intervals between such tests are not related

to conditions which cause accidents. The

proposed changes do not involve a change to

the plant design or operation. Therefore, this

change does not involve a significant

increase in the probability of any accident

previously evaluated.

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’

contributed to the technical bases for Option

B of 10 CFR 50 Appendix J. NUREG-1493

contains a detailed evaluation of the

expected leakage from containment and the

associated consequences. The increased risk

due to lengthening of the intervals between

containment leakage tests was also evaluated

and found acceptable. Using a statistical

approach, NUREG-1493 determined the

increase in the expected dose to the public

from extending the testing frequency is

extremely small. It also concluded that a

small increase is justifiable due to the

benefits which accrue from the interval

extension. The primary benefit is in the

reduction in occupational exposure. The

reduction in the occupational exposure is a

real reduction, while the small increase to

the public is statistically derived using

conservative assumptions. Therefore, this

change does not involve a significant

increase in the consequences of any accident

previously evaluated.

The proposed change does not involve

modifications to any existing equipment. The

proposed change will not affect the operation

of the plant or the manner in which the plant

is operated. The reduced testing frequency

will not affect the testing methodology.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not change the

performance methodology of the containment

leakage rate testing program. However, the

proposed change does affect the frequency of

containment leakage rate testing. With an

increased frequency between tests, the

proposed change does increase the

probability that a increase in leakage could

go undetected for a longer period of time.

Operational experience has demonstrated the

leak tightness of the containment buildings

has been significantly below the allowable

leakage limit.

The margin of safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to containment leakage rates. The

limitation on containment leakage rate is

designed to ensure the total leakage volume

will not exceed the value assumed in our

accident analysis. The margin of safety for

the offsite dose consequences of postulated

accidents directly related to containment

leakage is maintained by meeting the 1.0 La

acceptance criteria. The proposed change

maintains the 1.0 La acceptance criteria.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

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

NRC Project Director: William D.

Beckner

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:

September 19, 1996

Description of amendment request:

The proposed changes to Plant Hatch

Units 1 and 2 Technical Specifications

would revise the Surveillance

Requirements (SRs) addressing the

reactor vessel pressure and temperature

(P/T) limits. The affected SRs are

3.4.9.1, 3.4.9.2, 3.4.9.3, 3.4.9.4, 3.4.9.5,

3.4.9.6, and 3.4.9.7, and the

corresponding Units 1 and 2 Figures

3.4.9-1, 3.4.9-2, and 3.4.9-3, which show

P/T limit curves for inservice leak and

hydrostatic testing, non-nuclear heatup

and cooldown, and criticality,

respectively.

The P/T curves would be changed to

allow separate monitoring of the three

major regions of the reactor pressure

vessel (RPV) (i.e., the upper vessel and

flange region, the beltline region, and

the bottom head region), and to extend

the validity of the Unit 1 curves to 32
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 129

Effective Full Power Years (EFPY).

Separate monitoring would alleviate the

difficulties with meeting certain

temperature requirements due to the

artificial limits imposed by the current

P/T curves.

In support of the proposed changes,

General Electric (GE) prepared and

issued GENE-523-A137-1295, ‘‘E. I.

Hatch Nuclear Power Station, P-T Curve

Modification for Unit 1 and Unit 2,’’

which is provided in the submittal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Pressure and temperature (P/T) limits for

the reactor pressure vessel (RPV) are

established to ensure brittle fracture of the

vessel does not occur.

A. The proposed changes merely clarify the

Applicability of the P/T limits for each of the

low pressure conditions by replacing the

word ‘‘performed’’ with ‘‘met’’, adding Notes

to Surveillance Requirements, incorporating

the requirements of Notes into the

Surveillance Requirements, and modifying

the Frequency statements. Conditions 2, 3,

and 4, discussed in Enclosure 1 ‘‘justification

of changes’’, [of the licensee’s application]

have their own Surveillance Requirements.

Temperature requirements for Condition 1

are specified in the Bases. This proposed

change only clarifies which Surveillance

Requirement applies to each operating

configuration. No reduction in Surveillance

Frequencies is proposed.

B. The proposed revisions to the operating

limits curves for inservice leak and

hydrostatic testing, and the heatup and

cooldown allow independent monitoring of

the three RPV regions; i.e., the bottom head,

the upper vessel and flange, and the core

beltline. The three Unit 1 curves, including

the criticality curve, were extended to 32

Effective Full Power Years (EFPY), and a

correction to the Unit 1 criticality curve was

made. Operating limits for each of the curves

were evaluated in accordance with the

methodology given in the applicable ASME

Codes; Regulatory Guide 1.99, Rev. 2, and

Appendix G of 10 CFR [Part] 50.

The actual limits in the inservice leak and

hydrostatic testing curves, and the heatup

and cooldown curves were not relaxed.

Therefore, segregating the curves into the

three affected vessel regions does not

represent a reduction in the actual P/T

requirements. The current P/T curves

represent a composite of the three regions,

with each point representing the limiting

region. Regions of the vessel that are not

limiting at a specific point are, therefore,

artificially restrained. Upon implementation

of the proposed changes, each vessel region

will have its own curve, with its own true

limit.

Since the proposed changes do not affect

the recirculation piping, the probability and

the consequences of a loss of coolant

accident are not increased. Likewise, no

other previously evaluated accidents or

transients, as defined in Chapters 14 and 15

of the Units 1 and 2 Final Safety Analysis

Reports, are affected by the proposed

changes.

In summary, the proposed changes do not

represent a relaxation of any actual operating

limit and do not reduce the Frequency of any

Surveillance. Three of the four operating

configurations of the RPV are covered by

Surveillance Requirements. Temperature

limitations for the head removed from the

vessel are given in the Bases. The operating

limits were developed using the approved

methodology contained in 10 CFR [Part] 50,

Appendix G. Therefore, the probability and

consequences of a brittle fracture of the RPV

are not increased.

2. Do the proposed changes create the

possibility of a new or different type of

accident from any previously evaluated.

Implementing the low pressure changes, or

the new operating limit curves, does not alter

the design or operation of any system

designed for the prevention or mitigation of

accidents. The proposed changes do not

introduce any new type of normal or

abnormal operating mode or failure mode.

All P/T limits for the Unit 1 and the Unit 2

reactor vessels continue to be monitored per

the requirements of 10 CFR [Part] 50,

Appendices G and H. Therefore, the

proposed changes do not create the

possibility of a new type of accident.

3. Do the proposed changes involve a

significant reduction in the margin of safety?

The purpose of the P/T limits is to ensure

a brittle fracture of the RPV does not occur.

The proposed Technical Specifications

changes for the low pressure conditions are

made for clarification purposes. No operating

limits or Surveillance Requirements are

relaxed. The wording of current Technical

Specifications SRs 3.4.9.1, 3.4.9.2, 3.4.9.5,

3.4.9.6, and 3.4.9.7 could result in overly

conservative application of the requirements.

The proposed amendment is written to

remove the ambiguity in that the

Applicability and Frequency of each

Surveillance Requirement are clear. Neither

the acceptance criteria nor the Surveillance

Frequency of any Surveillance is reduced.

Furthermore, the four possible RPV

configurations are all adequately monitored.

As a result, the margin of safety for the low

pressure conditions is not significantly

reduced due to the proposed changes.

The Unit 1 operating curves were extended

to 32 EFPY using approved methodologies.

More operational margin is provided,

because the three vessel regions (upper vessel

and flange, beltline, and bottom head) are

being separated for the inservice leak and

hydrostatic testing curve, and the heatup and

cooldown curve. Although this separation

results in more operating margin for certain

vessel regions, it does not represent a

significant reduction in the margin of safety.

As described previously, the current

Technical Specifications curves represent a

composite of the three regions. Thus, the

curves represent the temperature for the

limiting region at a particular point. The

regions that are not limiting at a particular

point are artificially restricted. Separating the

three regions, as proposed, eliminates false

limits. The true limit for each region is

preserved and uncompromised, based on the

use of approved methodologies.

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

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: October

7, 1996

Description of amendment request:

The proposed changes to Plant Hatch

Unit 1 and Unit 2 Technical

Specifications (TS) would revise

Surveillance Requirements (SR) 3.1.7.7

and 3.4.3.1, and Limiting Conditions for

Operation (LCO) 3.4.3, 3.5.1, and

3.6.1.6, to increase the nominal

mechanical pressure relief setpoints for

all of the 11 safety/relief valves (SRV) to

1150 psig and allow operation with one

SRV and its associated functions

inoperable. The proposed changes

would reduce the potential for SRV

pilot leakage and the potential for forced

outages due to an inoperable SRV

during a fuel cycle.

The existing TS require that during

continuous operation, all of the 11 SRVs

remain OPERABLE in the safety mode,

7 in the Automatic Depressurization

System (ADS) mode, and 4 in the LowLow Set (LLS) mode. If one SRV is

inoperable for longer than the duration

specified in the applicable Action

Statements, the plant must be placed in

a Cold Shutdown Condition. Analyses

have been completed which show that,

with one SRV out of service, all

transient/accident criteria can still be

met. Increasing the nominal mechanical

relief setpoints will increase the simmer

margin (i.e., the difference between the

SRV setpoints and the vessel steam

dome pressure), thereby potentially

reducing SRV pilot leakage which may

occur during a typical operating cycle.
130 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

As a result of increasing the mechanical

relief setpoints for the SRVs, the

Standby Liquid Control (SLC) System

pump test discharge pressure is

increased to 1232 psig. The High

Pressure Coolant Injection (HPCI) and

Reactor Core Isolation Cooling (RCIC)

systems are capable of operating at this

increased pressure.

In support of the proposed changes,

General Electric (GE) prepared NEDC32041P, ‘‘Safety Review for Edwin I.

Hatch Nuclear Power Plant Units 1 and

2 Updated Safety/Relief Valve

Performance Requirements,’’ Revision 2,

dated April 1996, which was included

in the submittal.

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 changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The SRVs serve to mitigate postulated

transients and accidents; the proposed

changes do not alter the function or mode of

operation of the SRVs. The probability of an

OPERABLE or an INOPERABLE SRV

inadvertently opening or failing to open or

close is not affected by these changes.

Therefore, the probability of an accident is

not increased. Analysis(a) has been performed

which considers the consequences of the

various transients and accidents with the

increased setpoints and with one SRV

inoperable. The analysis also considers the

impact on ECCS [Emergency Core Cooling

System] performance, including HPCI and

RCIC. The analysis has shown that the

consequences of an accident with the

increased SRV setpoints and with one SRV

inoperable are not increased.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

Revising the nominal SRV setpoint only

changes when the SRV opens in its

mechanical relief mode; the operation of the

SRV and any other existing equipment is not

altered. Operation with one SRV inoperable

was evaluated(a) and does not introduce any

new failure modes. The impact on the

operation and design of other systems and

components has been evaluated,(a) including

ECCS and SLC. No new operating modes or

failure modes are introduced. Thus, these

changes do not contribute to a new or

different type of accident.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

The change in SRV setpoint and operation

with one SRV inoperable was evaluated

relative to the applicable safety system

settings and found to remain acceptable. For

example, the proposed changes were

evaluated against peak clad temperature

limits, ECCS operation, ASME Code

overpressurization limits, the MINIMUM

CRITICAL POWER RATIO Safety Limit, and

containment design limits; no significant

reduction in the margin of safety was

identified(a)

.

(a) GE Report NEDC-32041P, ‘‘Safety

Review for Edwin I. Hatch Nuclear Power

Plant Units 1 and 2 Updated Safety/Relief

Valve Performance Requirements, Revision 2

(Proprietary), April 1996’’.

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

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: October

29, 1996

Description of amendment request:

The proposed amendments would

change the Technical Specifications

(TS) for Plant Hatch Units 1 and 2

associated with the installation of a

digital Power Range Neutron Monitoring

(PRNM) system and the incorporation of

long-term stability solution hardware.

In response to Generic Letter 94-02,

‘‘Thermal-Hydraulic Instabilities in

Boiling Water Reactors,’’ Georgia Power

Company (GPC) selected General

Electric (GE) Option III as the long-term

stability solution. Option III detects core

instabilities and provides a reactor

scram signal to the Reactor Protection

System (RPS). The long-term stability

solution, GE Option III, is supported by

the BWR Owners’ Group Topical Report

NEDO-31960-A submitted to the NRC

for approval in May 1991, and NEDO31960-A, Supplement 1, submitted to

the NRC for approval in March 1992.

The NRC issued a Safety Evaluation

Report (SER) for NEDO-31960-A and

Supplement 1 in July 1993. BWR

Owners’ Group Topical Report NEDO32465, submitted to the NRC in June

1995, provides additional analysis for

the detection and suppression

methodology (Option III).

To execute the stability solution

software, the Average Power Range

Monitor (APRM) and Rod Block Monitor

(RBM) electronics would be replaced

with a PRNM system based on digital

GE Nuclear Measurements Analysis and

Control NUMAC modules.

Implementation of the PRNM would

affect the RPS and Control Rod Block TS

3.3.1.1, 3.3.2.1, 3.4.1 and 3.10.8.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The purpose of the proposed amendment

is to incorporate the Power Range Neutron

Monitoring (PRNM) retrofit and Oscillation

Power Range Monitor (OPRM) installation.

The types of Average Power Range Monitor

(APRM) Functions that are credited to

mitigate accidents were previously evaluated.

The proposed OPRM Upscale Function is

implemented in the same hardware that

implements the APRM Functions. The

change to a two-out-of-four RPS [Reactor

Protection System] logic was analyzed and

determined to be equal to the original logic.

The modification involves equipment that

is intended to detect the symptoms of some

accidents and initiate mitigating action. The

worst case failure of the equipment involved

in the modification is a failure to initiate

mitigating action (scram), but no failure can

cause an accident. As discussed in the bases

for proposed changes, the PRNM replacement

system is designed to perform the same

operations as the existing Power Range

Monitoring (PRM) system and to meet or

exceed all of its operational requirements.

Therefore, it is concluded that the probability

of an accident previously evaluated is not

increased as a result of replacing the existing

equipment with the PRNM equipment.

* * * *

Human-machine interface (HMI) failures in

the current system could be related to

incorrectly adjusted settings, incorrect

reading of meters, and failure to return the

equipment to the normal operating

configuration. There are comparable failure

modes for some of these problems in the

digital system where an erroneous

potentiometer adjustment in the current

system is equivalent to an erroneous digital

entry in the replacement system. Certain

potential ‘‘failure to reconfigure errors’’ in

the current system have no counterpart in the

replacement system, because any

reconfiguration’’ is automatically returned to

normal by the system. Also, since parameters

are available for review at any time, even if

an error, such as a digital entry error occurs,

it is more likely that the error would be

almost immediately detected by recognition

that the displayed value is not the correct

one.

The failure analysis of the current system

assumes certain rates of human error. The

rates for the replacement system will be

lower and, hence, are bounded by the FSAR

[Final Safety Analysis Report] analysis.

Therefore, GPC [Georgia Power Company]

concludes the proposed changes do not
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 131

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The APRM Trip Functions credited in the

accident analyses are retained in the PRNM

retrofit. The response time of the new

electronics meets or exceeds the required

response criteria. No new interfaces or

interactions with other equipment will

introduce any new failure modes.

The modification involves equipment that

is intended to detect the symptoms of some

accidents and initiate mitigating action. The

worst-case failure of the equipment involved

in the modification is a failure to initiate

mitigating action (scram), but no failure can

cause an accident. This is unchanged from

the current system.

Software common-cause failures can at

most cause the system to fail to perform its

safety function. In that case, it could fail to

initiate action to mitigate the consequences

of an accident, but would not cause one.

The new system is a digital system with

software (firmware) control. As such, it has

‘‘central’’ processing points and software

controlled digital processing where the

current system had analog and discrete

component processing. The result is that the

specific failures of hardware and potentially

common-cause software failures are different

from the current system. Also, automatic selftest results in some cases in a direct trip as

a result of a hardware failure where the

current system may have remained ‘‘as-is’’.

However, when these are evaluated at the

system level, there are no new effects. In

general, FSARs assume simplistic failure

modes (relays for example) but do not

specifically evaluate such effects as self-test

detection and automatic trip or alarm.

The effects of software common-cause

failure are mitigated by hardware design and

system architecture. The replacement

equipment is fully qualified to operate in its

installed location and will not affect other

equipment.

Therefore, GPC concludes the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The replacement equipment provides the

same function as the original electronics.

Response time and operator information are

either maintained or improved. The

equipment was qualified, where appropriate,

to assure its intended safety function is

performed. The replacement system has

improved channel trip accuracy compared to

the current system and meets or exceeds

system requirements assumed in setpoint

analysis. The channel response time exceeds

the requirements. The channel indicated

accuracy is improved over the current

system, and meets or exceeds system

requirements. The replacement system meets

or exceeds all system requirements.

The BWROG [BWR Owners’ Group]

Stability Option III was developed to meet

the requirements of GDC [General Design

Criterion] 10 and GDC 12 by providing a

hardware system that detects the presence of

thermal-hydraulic instabilities and

automatically initiates the necessary actions

to suppress the oscillations prior to violating

the MCPR [maximum critical power ratio]

Safety Limit. The NRC has reviewed and

accepted the Option III methodology

described in Licensing Topical Report NEDO31960 and concluded this solution will

provide the intended protection. Therefore, it

is concluded that there will be no reduction

in the margin of safety as defined in the

Technical Specifications as a result of the

installation of the OPRM system and the

simultaneous removal of the operating

restrictions imposed by the ICAs [item

control areas].

Therefore, GPC concludes 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: 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

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 20, 1996

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

allow the Vice President to designate

the Safety Audit and Review Committee

(SARC) Chairperson, to change the work

hours limitation in accordance with

guidance in GL 82-12, ‘‘Nuclear Power

Plant Staff Working Hours;’’ to change

radioactive shipments record retention

requirements to comply with recent 10

CFR Part 20 changes; and other editorial

changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The changes requested are

administrative in nature. Paragraph 3.D was

placed in the License by Amendment No. 155

to authorize Omaha Public Power District

(OPPD) to increase the storage capacity of the

FCS spent fuel pool. Amendment No. 155

stated that the TS as issued would be

effective when the last new rack was

installed. Since the last new rack was

installed on

August 8, 1994, Paragraph 3.D is no

longer necessary and should

be deleted from the License.

Table of Contents, Section 6.0, ‘‘Interim

Special Technical Specifications,’’

Subsections 6.1 through 6.4 are proposed for

deletion because all of the Specifications

referred to have been deleted by previous

Amendments.

The revision proposed for TS 2.15 (Item 2C

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

insert the correct terminology (Pressurizer

Low/Low Pressure) into the Functional Unit

description.

The revision proposed for TS 5.2 will

require the control of overtime worked by

personnel to be in accordance with the NRC

Policy Statement on working hours (Generic

Letter 82-12) in lieu of stating the specific

times requirements from the Policy as the

current TS does. This option is in accordance

with NUREG-1432, Standard TS for

Combustion Engineering Plants,

Specification 5.2.2e, and will allow work

groups to be on twelve hour shifts.

The revision proposed for TS 5.5.2.2 will

replace the specific title of the Chairperson

of the Safety Audit and Review Committee

and replace it with ‘‘Member as appointed by

the Vice President.’’ This will allow the

flexibility to change chairmanship of the

committee amongst the members.

The revision to TS 5.10 concerning

retention of records of radioactive shipments

will update the TS to current 10 CFR 20

requirements. Plant procedures already

comply with current 10 CFR 20 record

retention requirements. The addition of the

Section 5.0 title corrects a minor format

discrepancy.

These proposed revisions are

administrative in nature. The proposed

revisions have no effect on any initial

assumptions or operating restrictions

assumed in any accident, nor do these

changes have any effect on equipment

required to mitigate the consequences of an

accident. Therefore the proposed revisions

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed revisions correct minor

errors, remove outdated information, are

consistent with changes in organizational

structure, 10 CFR Part 20, or NUREG-1432,

‘‘Combustion Engineering Standard

Technical Specifications (STS). These

changes will not result in any physical

alterations to the plant configuration,

changes to setpoint values, or changes to the

application of setpoints or limits. No new

operating modes are proposed as a result of

these changes. Therefore the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.
132 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

3. The proposed change does not involve

a significant reduction in a margin of safety.

The revisions listed above correct minor

errors, remove outdated information, or are

consistent with changes in organizational

structure, 10 CFR Part 20, or Standard TS.

These changes will not result in any physical

alterations to the plant configuration,

changes to setpoint values, or changes to the

application of setpoints or limits. Therefore

the proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: October

28, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section 3/

4.8.1, ‘‘A.C. Sources,’’ TS Section 3/

4.8.2, ‘‘Onsite Power Distribution

Systems,’’ TS Table 4.8.1, ‘‘Battery

Surveillance Requirements,’’ and the

associated bases. Surveillance

requirements would be modified to

account for the increase in the fuel

cycle, consistent with Generic Letter 91-

04, ‘‘Changes in Technical Specification

Surveillance Intervals to Accommodate

a 24-month Fuel Cycle,’’ dated April 2,

1991. Administrative changes are also

proposed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervals from 18 to 24

months for the A.C. Offsite Sources, the

Emergency Diesel Generators and the Station

Batteries or the proposed revision to remove

the ‘‘during shutdown’’ restriction for

conduct of the battery performance test.

Results of the review of historical 18

month surveillance data and maintenance

records support an increase in the

surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because no potential for a

significant increase in a failure rate of a

system or component was identified during

these reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

Initiating conditions and assumptions

remain as previously analyzed for accidents

in the DBNPS Updated Safety Analysis

Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

The proposed revision to reflect that the

battery charger performance test will

continue to be conducted on a[n] 18 month

surveillance interval is an administrative

change and does not affect previously

analyzed accidents.

The proposed revision to the Bases to

reflect that a change to a 24 month

surveillance test interval is an exception to

current guidance is an administrative change

and does not affect previously analyzed

accidents.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes and the assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these revisions

do not involve any physical changes to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated.

No changes are being proposed to the type

of testing currently being performed, only to

the length of the surveillance test interval

and to restrictions on conducting testing only

during shutdown conditions.

Results of the review of historical 18

month surveillance data and maintenance

records support an increase in the

surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because no potential for a

significant increase in a failure rate of a

system or component was identified during

these reviews.

The proposed revision to reflect that the

battery charger performance test will

continue to be conducted on a[n] 18 month

surveillance interval is an administrative

change and does not alter testing currently

being performed.

The proposed revision to the Bases to

reflect that a change to a 24 month

surveillance test interval is an exception to

current guidance is an administrative change

and does not alter testing currently being

performed.

3. Not involve a significant reduction in a

margin of safety because the results of the

historical 18 month surveillance data and

maintenance records review identified no

potential for a significant increase in a failure

rate of a system or component due to

increasing the surveillance test interval to 24

months. Existing system and component

redundancy is not being changed by these

proposed changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences, consequently there

are no significant reductions 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 Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Virginia Electric and Power Company,

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

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

November 26, 1996

Description of amendment request:

The proposed changes would eliminate

the records retention requirements from

the administrative section of the

Technical Specifications (TS) in

accordance with NRC Administrative

Letter95-06, ‘‘Relocation of Technical

Specifications Administrative Controls

Related to Quality Assurance.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Specifically, operation of the ... North

Anna Power [Station] in accordance with the

proposed Technical Specifications changes

will not:
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 133

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed

administrative changes do not affect

equipment or its operation. Therefore, the

likelihood that an accident will occur is

neither increased nor decreased by relocating

record retention requirements from the

Technical Specifications to the Operational

Quality Assurance Program. This TS change

will not impact the function or method of

operation of plant equipment. Thus, a

significant increase in the probability of a

previously analyzed accident does not result

due to this change. No systems, equipment,

or components are affected by the proposed

changes. Thus, the consequences of any

accident previously evaluated in the UFSAR

[Updated Final Safety Analysis Report] are

not increased by this change.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated. The proposed change

does not alter the design or operations of the

physical plant. Since record retention

requirements are administrative in nature, a

change to these requirements does not

contribute to accident initiation, an

administrative change related to this activity

does not produce a new accident scenario or

produce a new type of equipment

malfunction. [These] changes do not alter any

existing accident scenarios. The proposed

administrative change does not affect

equipment or its operation, and, thus, does

not create the possibility of a new or different

kind of accident. Therefore, the proposed

change does not create the possibility of a

new or different kind of accident.

(3) Involve a significant reduction in a

margin of safety. Section 6.0 of the North

Anna ... Technical Specifications does not

have a basis description. The proposed

administrative change does not affect

equipment or its operation, and, thus, does

not involve any reduction in the margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: Mark Reinhart,

Acting

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 3, 1996

Description of amendment request:

This amendment request proposes to

revise the technical specifications

associated with the inspection of the

reactor coolant flywheel to provide an

exception to the recommendations of

Regulatory Guide 1.14, Revision 1,

‘‘Reactor Coolant Pump Flywheel

Integrity.’’ The proposed exception

would allow either an ultrasonic

volumetric examination or surface

examination to be performed at

approximately 10-year intervals. In

addition, a correction of the issuance

date of a referenced regulatory guide is

included.

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 p presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The safety function of the RCP [reactor

coolant pump] flywheels is to provide a

coastdown period during which the RCPs

would continue to provide reactor coolant

flow to the reactor after loss of power to the

RCPs. The maximum loading on the RCP

flywheel results from overspeed following a

LOCA [loss-of-coolant accident]. The

maximum obtainable speed in the event of a

LOCA was predicted to be less than 1500

rpm. Therefore, a peak LOCA speed of 1500

rpm is used in the evaluation of RCP

flywheel integrity in WCAP-14535. This

integrity evaluation shows a very high flaw

tolerance for the flywheels. The proposed

change does not affect that evaluation.

Reduced coastdown times due to a single

failed flywheel is bounded by the locked

rotor analysis, therefore, it would not place

the plant in an unanalyzed condition.

Therefore, these changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed amendment does not create

the possibility of a new or different kind of

accident from any previously evaluated since

the proposed amendments will not change

the physical plant or the modes of plant

operation defined in the facility operating

license. No new failure mode is introduced

due to the proposed change, since the

proposed change does not involve the

addition or modification of equipment, nor

do they alter the design or operation of

affected plant systems, structures, or

components.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are basically

unchanged by the proposed amendment. The

results of the flywheel inspections performed

have identified no indications affecting

flywheel integrity. As identified in WCAP14535, detailed stress analysis as well as risk

analysis have been completed with the

results indicating that there would be no

change in the probability of failure for RCP

flywheels if all inspections were eliminated.

Therefore these changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 3, 1996

Description of amendment request:

This amendment request proposes to

correct the reference to the Action

Statement for Item 7.b, RWST Level -

Low-Low Coincident with Safety

Injection, Table 3.3-3, Engineered Safety

Features Actuation System

Instrumentation, from Action 16 to

Action 28.

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.

Changing the reference from Action

Statement 16 to Action Statement 28 for

Functional Unit 7.b. of Table 3.3-3 will

reduce the probability for an automatic

switchover from the RWST [refueling water

storage tank] to an empty containment sump

to occur, while an RWST level channel is

inoperable or is being tested with its bistable

tripped, should an inadvertent safety

injection signal occur concurrent with a

single failure of a second RWST level

channel. The design of these channels does

not allow for operation or testing in bypass,

so Action Statement 16 is not applicable.

Changing to Action Statement 28 will limit
134 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

the duration that a channel could be

inoperable or be in test with its bistable

bypassed. This change does not involve any

design changes or hardware modifications,

and does not introduce any new potential

accident initiating conditions. The increase

in allowed outage time for this item was

evaluated and the associated unavailability

and risk was shown to be equivalent to, or

less than, that of other functional units

evaluated in WCAP-10271, Supplement 2,

Revision 1. Therefore, this proposed change

does not increase the probability of any

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not result in

any hardware changes and does not result in

a change in the manner in which the ESFAS

[engineered safety features actuation system]

provides plant protection. This change does

not alter the functioning of the ESFAS.

Rather, the likelihood or probability of the

ESFAS functioning properly is affected as

described above. This change will not change

the method by which any safety-related

system performs its function. Therefore, this

proposed change will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

This proposed change will not result in a

significant reduction in the margin of safety

defined for any technical specification since

it does not alter the manner in which safety

limits, limiting safety system settings, or

limiting conditions for operation are

determined.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

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:

June 28, 1996

Brief description of amendments: The

amendment would modify the technical

specifications (TS) to increase the

minimum required amount of

anhydrous trisodium phosphate (TSP)

in the containment baskets. TSP is used

to ensure that following a postulated

design basis loss of coolant accident

(LOCA), the containment sump pH is

maintained greater than or equal to

seven.

Date of issuance: December 10, 1996

Effective date: December 10, 1996, to

be implemented within 45 days from

the date of issuance.

Amendment Nos.: Unit 1 - 110; Unit

2 - 102; Unit 3 - 82

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47962) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 10, 1996.No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

June 21, 1996

Brief description of amendments: The

amendments revise the term ‘‘lifting

loads’’ used in Technical Specification

3.9.6b.2, Manipulator Crane, to ‘‘lifting

force.’’ This revision will clarify that the

static loads associated with the lifting

tool, drive rod, and control rod weights

are not included in the lifting force

limit.

Date of issuance: December 12, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 171 and 153

Facility Operating License Nos. NPF9 and NPF-17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47977) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 12, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

February 22, 1996, and as supplemented

by letters dated July 4 and September

20, 1996

Brief description of amendment: The

amendment revises Clinton Power

Station Technical Specification 3.3.4.1,

‘‘End of Cycle Recirculation Pump Trip

(EOC-RPT) Instrumentation,’’ by

deleting Surveillance Requirement

3.3.4.1.6 which requires the RPT breaker

interruption time to be determined at

least once per 60 months.

Date of issuance: December 13, 1996

Effective date: December 13, 1996

Amendment No.: 111
Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices 135

Facility Operating License No. NPF62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18169)

The supplemental letters of July 4 and

September 20, 1996, provided clarifying

information and did not include

significant changes relative to the

original Federal Register notice.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 13, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Niagara Mohawk Power Corporation,

Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of application for amendment:

July 12, 1996, as supplementedOctober

30, 1996.

Brief description of amendment: The

amendment revises TS 6.2.2.h regarding

the administrative controls for the

normal working hours of unit staff who

perform safety-related functions, and TS

6.2.2.i regarding an organizational

change. The changes authorize (1)

establishment of unit staff work

schedules that average 40 hours per

week using shifts as long as 12 hours,

and (2) elimination of the positions of

General Supervisor Operations and

Supervisor Operations.

Date of issuance: December 12, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 158

Facility Operating License No. DPR63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR 42280)

The October 30, 1996, letter provided

supplemental information that 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 December 12, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

July 12, 1996

Brief description of amendment: The

amendment revises Technical

Specification Section 6.2.2.i regarding

the administrative controls for the

normal working hours of unit staff who

perform safety-related functions. The

change allows the establishment of unit

staff work schedules that average 40

hours per week using shifts as long as

12 hours.

Date of issuance: December 12, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 78

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR 42281)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 12, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

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:

July 28, 1995, as supplemented October

25, 1995, and August 9, 1996

Brief description of amendments: The

amendments revise the 250 volt DC

profiles in the Technical Specifications

for the two units to reflect new load

profile calculations.

Date of issuance: December 17, 1996

Effective date: Unit 1, as of date of

issuance, to be implemented within 30

days; Unit 2, as of date of issuance, to

be implemented prior to Startup

following the Eighth Refueling and

Inspection Outage for Unit 2, which is

scheduled for the Spring of 1997.

Amendment Nos.: 162 and 133

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 13, 1995 (60 FR

47622) The supplemental letters

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination nor the Federal Register

notice.The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 17, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Power Authority of the State of New

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

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

June 12, 1992, as supplemented

September 17, 1992, March 17, 1993,

August 17, 1993, August 18, 1993,

December 29, 1993, June 29, 1995,

August 15, 1996, October 3,

1996,October 23, 1996, November 14,

1996, November 20, 1996 (JPN-96-045),

November 20, 1996 (JPN-96-046), and

November 27, 1996.

Brief description of amendment: The

amendment modifies

Facility Operating License No. DPR-59

and the James A. FitzPatrick Nuclear

Power Plant (JAFNPP) Technical

Specifications (TSs) to authorize an

increase in the maximum power level of

JAFNPP from 2436 MWt to 2536 MWt.

The amendment also approves changes

to the TSs to implement uprated power

operation.

Date of issuance: December 6, 1996

Effective date:

As of the date of issuance to be

implemented upon plant startup

following the refueling outage cycle 13.

Amendment No.: 239

Facility Operating License No. DPR59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 2, 1994 (59 FR 4943)

The letters dated September 17, 1992,

March 17, 1993, August 17, 1993,

August 18, 1993, December 29, 1993,

June 29, 1995, August 15, 1996,October

3, 1996, October 23, 1996, November 14,

1996, November 20, 1996, (JPN-96-045),

November 20, 1996, (JPN-96-046), and

November 27, 1996, provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 6, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.
136 Federal Register / Vol. 62, No. 1 / Thursday, January 2, 1997 / Notices

1 15 U.S.C. 78s(b)(1).

2 17 CFR 240.19b–4.

3On December 17, 1996, the Exchange filed

Amendment No. 1 to the proposed rule change.

Amendment No. 1 is a technical amendment,

correcting Exhibit I, Section I to the filing. See letter

from Margaret Abrams, Senior Attorney, CBOE to

Janice Mitnick, Attorney, Division of Market

Regulation, SEC, dated December 17, 1996.

Public Service Electric & Gas Company,

Docket No. 50-311, Salem Nuclear

Generating Station, Unit No. 2, Salem

County, New Jersey Date of application

for amendment: September 20, 1996, as

supplemented September 30, 1996

Brief description of amendment: The

amendment changes Technical

Specification Surveillance Requirement

4.7.7.b.4 for the Auxiliary Building

Exhaust Air Filtration System, and its

associated Bases, to indicate that the

specified flowrate applies only to

system testing.

Date of issuance: December 12, 1996

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment No. 168

Facility Operating License No. DPR75: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 23, 1996 (61 FR

55040) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

December 12, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

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:

August 27, 1996, as supplemented

October 24, 1996

Brief description of amendments: The

amendment to Unit 2 deletes License

Condition 2.C.(24)(a) which required

establishment by June 3, 1981, of

regularly scheduled 8-hour shifts

without reliance on routine use of

overtime. The amendments to both

Units 1 and 2 revise Technical

Specification 6.2.2 to delete the

reference to Generic Letter 82-12,

‘‘Nuclear Plant Staff Working Hours,’’

and require that administrative controls

be established which will ensure that

adequate shift coverage is maintained

without heavy use of overtime for

individuals.

Date of issuance: December 17, 1996

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos. 186 and 169

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

revised the Technical Specifications for

both units and License for Unit 2 only.

Date of initial notice in Federal

Register: September 12, 1996 (61 FR

48175) The October 24, 1996, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination or the original notice.The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated December 17, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

May 29, 1996

Brief description of amendments:

These amendments revise Technical

Specification (TS) Surveillance

Requirement 3.5.1.4 to increase the

minimum boron concentration in the

safety injections tanks from 1850 ppm to

2200 ppm.

Date of issuance: December 6, 1996

Effective date: December 6, 1996, to

be implemented within 30 days from

the date of issuance.

Amendment Nos.: Unit 2 - 135; Unit

3 - 124

Facility Operating License Nos. NPF10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 31, 1996 (61 FR 40029)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated December 6, 1996. No

significant hazards consideration

comments received: No.Temporary

Local Public Document Room

location: Science Library, University of

California, P. O. Box 19557, Irvine,

California 92713

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

September 27, 1996, as supplemented

on October 25, and November 18, 1996

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant Technical Specification

requirements related to the low

temperature overpressure protection

(LTOP) system. Specifically, the LTOP

curve is modified to define 10 CFR Part

50, Appendix G pressure temperature

limitations for LTOP evaluation through

the end of operating cycle (EOC) 33. In

addition, the LTOP enabling

temperature and the temperature

required for starting a reactor coolant

pump have been changed consistent

with the design basis for the LTOP

system. Finally, the TS bases were

changed consistent with the changes

described above.

Date of issuance: December 13, 1996

Effective date: December 13, 1996, to

be implemented within 30 days.

Amendment No.: 130

Facility Operating License No. DPR43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 7, 1996 (61 FR 52472)

The October 25 and November 18, 1996,

submittals provided supplemental

information that did not change the

initial proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 13, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Dated at Rockville, Maryland, this 24th day

of December 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation

[Doc. 96-33254 Filed 12-31-96; 8:45 am]

BILLING CODE 7590-01-F

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 34–38086; File No. SR–CBOE–

96–69]

Self-Regulatory Organizations; Notice

of Filing and Order Granting

Accelerated Approval of Proposed

Rule Change by the Chicago Board

Options Exchange, Incorporated

Relating to Calculating Blue Sheets

Violation Aggregate Fines on a Rolling

Year Basis

December 26, 1996.

Pursuant to Section 19(b)(1) of the

Securities Exchange Act of 1934

(‘‘Act’’) 1, and Rule 19b–4 2 thereunder,

notice is hereby given that on November

20, 1996,3 the Chicago Board Options

Exchange, Incorporated (‘‘CBOE’’ or

‘‘Exchange’’) filed with the Securities

and Exchange Commission (‘‘SEC’’ or