Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42597

Dated: August 10, 1995.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 95–20236 Filed 8–15–95; 8:45 am]

BILLING CODE 7590–01–M

Biweekly Notice

Applications and Amendments to

Facility Operating LicensesInvolving

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 July 21,

1995, through August 4, 1995. The last

biweekly notice was published on

Wednesday, August 2, 1995 (60 FR

39430).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m.

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By September 15, 1995, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if
42598 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

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, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

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: March

15, 1995, as supplemented on June 29,

1995.

Description of amendments request:

The proposed amendments would

revise the Calvert Cliffs Nuclear Power

Plant, Units Nos. 1 and 2, Technical

Specifications (TSs) Section 6,

‘‘Administrative Controls,’’ to be

consistent with the guidance provided

in NUREG-1432, ‘‘Standard Technical

Specifications, Combustion Engineering

Plants.’’ The proposed changes will

relocate several requirements to other

documents and programs consistent

with NUREG-1432 and other NRC

guidance addressing the administrative

section of the TSs such as the ‘‘Final

Policy Statement on Technical

Specification Improvements for Nuclear

Power Reactors,’’ published in the

Federal Register on July 22, 1993 (58 FR

39132).

The Commission indicated that

compliance with the Final Policy

Statement satisfies Section 182a of the

Act. In particular, the Commission

indicated that certain items could be

relocated from the TSs to licenseecontrolled documents, consistent with

the standard enunciated in Portland

General Electric Co. (Trojan Nuclear

Plant), ALAB-531, 9 NRC 263, 273

(1979). In that case, the Atomic Safety

and Licensing Appeal Board indicated

that ‘‘technical specifications are to be

reserved for those matters as to which

the imposition of rigid conditions or

limitations upon reactor operation is

deemed necessary to obviate the

possibility of an abnormal situation or

event giving rise to an immediate threat

to the public health and safety.’’ The

policy statement encouraged licensees

to adopt the applicable improved STSs

and provided some guidance for the

conversion from the present plantspecific TSs to the improved Standard

TSs.

The proposed changes will provide

significant human factors improvement

to the TSs by accomplishing the

following: (1) relocating existing

requirements to licensee controlled

documents consistent with the policy

statement; (2) eliminating requirements

which duplicate regulations; (3)

relocating similar requirements within

the same section; (4) editorial changes;

and (5) adding requirements consistent

with NUREG-1432.

In addition, the licensee proposes

dual rolls for the Shift Technical

Advisor (STA) and the establishment of

a TS Bases Control Program. Allowing

the STA to perform dual rolls is not

permitted by the current TSs, but the

current NRC guidance allows the STA to

perform a dual roll. The proposed new

TS Bases Control Program will define

the appropriate methods and reviews

required to implement a TS Bases

change which is also consistent with the

current NRC guidance. Two other

proposed changes, not specifically

covered by the above groupings, include

a reduction in reporting requirements

and utilizing a more effective option for

estimating doses.

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.

Relocating existing requirements to

Baltimore Gas and Electric Company (BGE)-

controlled documents, eliminating

requirements which duplicate regulations,

locating similar requirements within the

same sections and making necessary editorial

corrections to incorporate the proposed

changes provide Technical Specifications

which are easier to use. Because existing

requirements are relocated to established

BGE programs where changes to those

programs are controlled by regulatory

requirements, there is no reduction in

commitment and adequate control is still

maintained. Likewise, the elimination of

requirements which duplicate regulations

enhances the usability of the Technical

Specifications without reducing

commitments. Locating similar requirements

within the same sections and making

necessary editorial corrections to incorporate

the proposed changes neither add nor delete

requirements, but merely clarify and improve

the readability and understanding of the

Technical Specifications. Since the

requirements remain the same, these changes

only affect the method of presentation and

would not affect possible initiating events for

accidents previously evaluated or any system

functional requirement. Therefore, the

proposed changes would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42599

Since the Shift Technical Adviser (STA) is

not considered an initiator to any previously

evaluated accident nor considered in the

accident’s response, the use of a dual role

STA would not increase the probability or

consequences of any previously evaluated

accident.

The Technical Specification Bases Control

Program provides controls which ensure

appropriate reviews of changes to the Bases.

Because NRC approval is still needed for

changes to the Bases which affect the

Technical Specifications, the proposed

Program would not affect the probability or

consequences of a previously evaluated

accident.

Eliminating the requirement for submitting

two reports which place unwarranted

administrative burden on both Baltimore Gas

and Electric Company and the NRC has no

affect on the probability or consequences of

an accident previously evaluated. Therefore,

the proposed changes would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Replacing the film badge with the

electronic personal dosimeter provides a

more effective, efficient, state-of-the art

option for estimating dose and would not

impact accidents previously evaluated.

Therefore, the proposed change would 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.

As discussed previously, relocating

existing requirements to BGE-controlled

documents, eliminating requirements which

duplicate regulations, locating similar

requirements within the same sections and

making necessary editorial corrections to

incorporate the proposed changes will not

affect any plant system or structure, nor will

it affect any system functional or operability

requirements. Consequently, no new failure

modes are introduced as a result of the

proposed changes. Therefore, these types of

changes would not create the possibility of a

new or different type of accident from any

accident previously evaluated.

Because the STA does not perform

equipment design or equipment

manipulation, the use of a dual role STA

would not create the possibility of a new or

different type of accident from any accident

previously evaluated. Since the Technical

Specification Bases Control Program

represents an administrative function

performed under existing regulatory controls,

it too would not create the possibility of a

new or different type of accident from any

previously evaluated.

The addition of new programs which

incorporate existing Technical Specification

requirements and commitments will have no

effect on the design or operation of the plant

and would not create the possibility of a new

or different type of accident from any

previously evaluated.

A reporting function such as report

submittals would not change the

configuration or operation of the plant.

Consequently, the elimination of the

requirement to submit the Startup Report and

the Special Report dealing with iodine

activity levels, would not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

Since the operation or configuration of the

plant is not changed by the type of personal

dosimeter, this change would not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

Therefore, the proposed changes would 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.

Relocating existing requirements to BGEcontrolled documents, eliminating

requirements which duplicate regulations,

locating similar requirements within the

same sections and making necessary editorial

corrections to incorporate the proposed

changes would not affect the Updated Final

Safety Analysis Report design bases, accident

analysis assumptions or any margin of safety

described in the Technical Specification

Bases. In addition, these proposed changes

do not affect effluent release limits,

monitoring equipment or practices.

Therefore, these proposed changes would not

involve a significant reduction in a margin of

safety.

The use of an STA should provide an

additional margin of safety in the accident

response function of licensed operators

beyond that considered in the accident

analysis. Since the STA is required to have

the same training and educational

qualifications in either the individual or dual

role, the use of a dual role STA should have

minimal impact. Consequently, the proposed

change would not involve a significant

reduction in a margin of safety. The

Technical Specification Bases Control

Program is an administrative change

controlling how Technical Specification

basis information is reviewed and

incorporated. Therefore, this change would

not involve a significant reduction in a

margin of safety.

The addition of new programs which

incorporate existing Technical Specification

requirements and commitments will have no

effect on the design or operation of the plant

and would not result in a significant

reduction in the margin of safety.

Activities described in the Startup Report

will continue to be performed and corrective

action taken when required. Similarly, iodine

activity levels will continue to be monitored

and actions taken, including the issuance of

a Licensee Event Report when conditions

warrant. Considering the above, elimination

of the two reporting requirements would

have no impact on the margin of safety.

Plant operating parameters are not affected

by the type of personnel monitoring device

used and as a consequence, would not

impact a margin of safety. Since the

replacement dosimeter provides a more

effective mechanism for estimating dose,

there is no degradation in personal safety

levels. Consequently, the proposed change

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the 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: Ledyard B.

Marsh

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendment

requests: September 17, 1993, as

supplemented July 28, 1995

Description of amendment requests:

As a result of findings by a Diagnostic

Evaluation Team inspection performed

by the NRC staff at the Dresden Nuclear

Power Station in 1987, Commonwealth

Edison Company (ComEd, the licensee)

made a decision that both the Dresden

Nuclear Power Station and sister site

Quad Cities Nuclear Power Station

needed attention focused on the existing

custom Technical Specifications (TS)

used.

The licensee made the decision to

initiate a Technical Specification

Upgrade Program (TSUP) for both

Dresden and Quad Cities. The licensee

evaluated the current TS for both

Dresden and Quad Cities against the

Standard Technical Specifications (STS)

contained in NUREG-0123, ‘‘Standard

Technical Specifications General

Electric Plants BWR/4.’’ The licensee’s

evaluation identified numerous

potential improvements such as

clarifying requirements, changing TS to

make them more understandable and to

eliminate interpretation, and deleting

requirements that are no longer

considered current with industry

practice. As a result of the evaluation,

ComEd has elected to upgrade both the

Dresden and Quad Cities TS to the STS

contained in NUREG-0123.

The TSUP for Dresden and Quad

Cities is not a complete adoption of the

STS. The TSUP focuses on (1)

integrating additional information such

as equipment operability requirements

during shutdown conditions, (2)

clarifying requirements such as limiting

conditions for operation and action
42600 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TS based on

the licensee’s responses to Generic

Letters (GL), and (4) relocating specific

items to more appropriate TS locations.

The September 17, 1993, and July 28,

1995, applications proposed to upgrade

only Section 3/4.5 (Emergency Core

Cooling Systems) of the Dresden and

Quad Cities TS.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Implementation

of these changes will provide increased

reliability of equipment assumed to operate

in the current safety analysis, or provide

continued assurance that specified

parameters remain within their acceptance

limits, and as such, will not significantly

increase the probability or consequences of a

previously evaluated accident.

Some of the proposed changes represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. The proposed amendment

for Dresden and Quad Cities Station’s

Technical Specification Section 3/4.5 are

based on STS guidelines or later operating

BWR plants’ NRC accepted changes. Any

deviations from STS requirements do not

significantly increase the probability or

consequences of any previously evaluated

accidents for Dresden or Quad Cities

Stations. The proposed amendment is

consistent with the current safety analyses

and has been previously determined to

represent sufficient requirements for the

assurance and reliability of equipment

assumed to operate in the safety analysis, or

provide continued assurance that specified

parameters remain within their acceptance

limits. As such, these changes will not

significantly increase the probability or

consequences of a previously evaluated

accident.

The associated systems that make up the

Emergency Core Cooling Systems are not

assumed in any safety analysis to initiate any

accident sequence for Dresden or Quad Cities

Stations; therefore, the probability of any

accident previously evaluated is not

increased by the proposed amendment. In

addition, the proposed surveillance

requirements for the proposed amendments

to these systems are generally more

prescriptive than the current requirements

specified within the Technical

Specifications. The additional surveillance

requirements improve the reliability and

availability of all affected systems and

therefore, reduce the consequences of any

accident previously evaluated as the

probability of the systems outlined within

Section 3/4.5 of the proposed Technical

Specifications performing their intended

function is increased by the additional

surveillances.

Create the possibility of a new or different

kind of accident from any previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, the

addition of requirements which are based on

the current safety analysis, and some minor

curtailments of the current requirements

which are based on generic guidance or

previously approved provisions for other

stations. These changes do not involve

revisions to the design of the station. Some

of the changes may involve revision in the

operation of the station; however, these

provide additional restrictions which are in

accordance with the current safety analysis,

or are to provide for additional testing or

surveillances which will not introduce new

failure mechanisms beyond those already

considered in the current safety analyses.

The proposed amendment for Dresden and

Quad Cities Station’s Technical Specification

Section 3/4.5 is based on STS guidelines or

later operating BWR plants’ NRC accepted

changes. The proposed amendment has been

reviewed for acceptability at the Dresden and

Quad Cities Nuclear Power Stations

considering similarity of system or

component design versus the STS or later

operating BWRs. Any deviations from STS

requirements do not create the possibility of

a new or different kind of accident

previously evaluated for Dresden or Quad

Cities Stations. No new modes of operation

are introduced by the proposed changes.

Surveillance requirements are changed to

reflect improvements in technique, frequency

of performance or operating experience at

later plants. Proposed changes to action

statements in many places add requirements

that are not in the present technical

specifications. The proposed changes

maintain at least the present level of

operability. Therefore, the proposed changes

do not create the possibility of a new or

different kind of accident from any

previously evaluated.

The associated systems that make up the

Emergency Core Cooling Systems are not

assumed in any safety analysis to initiate any

accident sequence for Dresden or Quad Cities

Stations. In addition, the proposed

surveillance requirements for affected

systems associated with the Emergency Core

Cooling Systems are generally more

prescriptive than the current requirements

specified within the Technical

Specifications; therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from any

previously evaluated.

Involve a significant reduction in the

margin of safety because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, the

addition of requirements which are based on

the current safety analysis, and some minor

curtailments of the current requirements

which are based on generic guidance or

previously approved provisions for other

stations. Some of the latter individual items

may introduce minor reductions in the

margin of safety when compared to the

current requirements. However, other

individual changes are the adoption of new

requirements which will provide significant

enhancement of the reliability of the

equipment assumed to operate in the safety

analysis, or provide enhanced assurance that

specified parameters remain with their

acceptance limits. These enhancements

compensate for the individual minor

reductions, such that taken together, the

proposed changes will not significantly

reduce the margin of safety.

The proposed amendment to Technical

Specification Section 3/4.5 implements

present requirements, or the intent of present

requirements in accordance with the

guidelines set forth in the STS. Any

deviations from STS requirements do not

significantly reduce the margin of safety for

Dresden or Quad Cities Stations. The

proposed changes are intended to improve

readability, usability, and the understanding

of technical specification requirements while

maintaining acceptable levels of safe

operation. The proposed changes have been

evaluated and found to be acceptable for use

at Dresden or Quad Cities based on system

design, safety analysis requirements and

operational performance. Since the proposed

changes are based on NRC accepted

provisions at other operating plants that are

applicable at Dresden or Quad Cities and

maintain necessary levels of system or

component reliability, the proposed changes

do not involve a significant reduction in the

margin of safety.

The proposed amendment for Dresden and

Quad Cities Stations will not reduce the

availability of systems associated with the

Emergency Core Cooling Systems when

required to mitigate accident conditions;

therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: for Dresden, Morris Public

Library, 604 Liberty Street, Morris,

Illinois 60450; for Quad Cities, Dixon

Public Library, 221 Hennepin Avenue,

Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690

NRC Project Director: Robert A. Capra
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42601

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: June 17,

1993, as supplemented July 5, 1995

Description of amendment request:

The initial proposed amendment

request dated June 17, 1993, was

previously noticed in the Federal

Register on July 21, 1993 (58 FR 39048).

The proposed amendment would revise

Technical Specification 5.3.1, ‘‘Fuel

Assemblies’’ to provide flexibility in the

repair of fuel assemblies containing

damaged and leaking fuel rods by

reconstituting the assemblies in

accordance with the guidance in

Generic Letter (GL) 90-02, Supplement

1, ‘‘Alternative Requirements For Fuel

Assemblies In The Design Features

Section Of Technical Specifications,’’

issued on July 31, 1992. The application

is also generally consistent with the

format and content of the improved

Standard Technical Specifications for

Westinghouse plants provided in

NUREG-1431.

Additional information was submitted

on July 5, 1995, that added TS changes

to increase the fuel enrichment limit

from 4.0 to 5.0 weight percent U-235

that were not previously included the

initial June 17, 1993, amendment

application. This additional information

is being noticed to provide for public

comment and opportunity for hearing.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee provided its analysis of the

issue of no significant hazards

consideration (58 FR 39048). The NRC

staff’s analysis of the July 5, 1995,

supplement against the standards of 10

CFR 50.92(c) is presented below.

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

There is no increase in the probability

or consequences of an accident in the

new fuel vault since the only accident

that would be affected by this change

would be a criticality accident and it

has been shown that the worst-case keff

under optimum moderation conditions

continues to be less than or equal to

0.98.

There is no increase in the probability

of a fuel drop accident in the Spent Fuel

Storage Pool since the mass of an

assembly will not be significantly

affected by the increase in fuel

enrichment. The likelihood of other

accidents, previously evaluated and

described in Section 9.1.2 of the Final

Safety Analysis Report (FSAR), is also

not affected by the proposed changes.

Since the increase in fuel enrichment

will allow for extended fuel cycles, it

could be postulated that there may be a

decrease in fuel movement and the

probability of an accident may likewise

be decreased. There is also no increase

in the consequences of a fuel drop

accident in the Spent Fuel Pool since

the fission product inventory of

individual fuel assemblies will not

change significantly as a result of

increased initial enrichment. In

addition, no change to safety-related

systems is being made.

Therefore, the consequences of a fuel

rupture accident remain unchanged. In

addition, it has been shown that keff is

less than or equal to 0.95, under all

conditions. Therefore, the consequences

of a criticality accident in the Spent

Fuel Pool remain unchanged as well.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not create

the possibility of a new or different kind

of accident since fuel handling

accidents (fuel drop and misplacement)

are not new or different kinds of

accidents. Fuel handling accidents are

already discussed in the FSAR for fuel

with enrichments up to 4.0 weight %

and additional analyses have been

performed for fuel with enrichment up

to 5.00 weight %.

3.

The proposed changes do not involve

a significant reduction in the margin of

safety.

The proposed change does not

involve a significant reduction in the

margin of safety since, in all cases, a

spent fuel pool keff less than or equal to

0.95 is being maintained. Criticality

analyses have also been performed that

show that the new fuel storage vault

will remain subcritical under a variety

of moderation conditions, from fully

flooded to optimum moderation. As

discussed above, the Spent Fuel Pool

will remain sufficiently subcritical

during any fuel misplacement accident.

Based on this analysis, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

supplemental amendment submittal

involves no significant hazards

consideration.

Local Public Document Room

location:: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Duke Power Company, Docket Nos. 50-

269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3,

Oconee County, South Carolina

Date of amendment request: July 26,

1995

Description of amendment request:

The proposed amendments would

provide a one-time extension of the

allowable outage time from 72 hours to

7 days. This extension is necessary to

implement a modification to the

degraded grid protection system and the

external grid trouble protection system.

Basis for proposed no significant

hazards considerationdetermination: 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:

Duke Power Company (Duke) has made

the determination that this amendment

request involves a No Significant

Hazards Consideration by applying the

standards established by NRC

regulations in 10 CFR 50.92. This

ensures that operation of the facility in

accordance with the proposed

amendment would not:(1) Involve a

significant increase in the probability

or consequences of an accident

previously evaluated:

Each accident analysis addressed within

the Oconee Final Safety Analysis Report

(FSAR) has been examined with respect to

the change proposed within this amendment

request. The design basis of the auxiliary

electrical systems is to supply the required

engineered safeguards (ES) loads of one unit

and the safe shutdown loads of the other two

units. The systems are arranged so that no

single failure will jeopardize plant safety.

The probability of any Design Basis

Accident (DBA) is not significantly increased

by this change. In addition, the consequences

of the accidents are within the bounds of the

FSAR analyses. The reliability of the

emergency power system is not significantly

affected by a one time extension of allowable

outage time for the overhead power path. The

underground power path is adequate to

assure operability of the Oconee ES loads.

Finally, the enhancement of the Degraded

[Grid] Protection System will eliminate a

concern which was expressed by the EDSFI

audit team.

(2) Create the possibility of a new or

different kind of accident from any kind of

accident previously evaluated:

Inoperability of the yellow bus is

functionally equivalent to inoperability of the

Keowee Main Step-up Transformer in that it

renders the overhead emergency power path

inoperable. The Keowee Main Step-up

Transformer is allowed to be inoperable for

a period not to exceed 28 days. This

Technical Specification requirement for the
42602 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

Keowee Main Step-up Transformer has been

reviewed and approved by the NRC.

Therefore, operation of ONS [Oconee Nuclear

Station] in accordance with this Technical

Specification amendment will not create any

failure modes not bounded by previously

evaluated accidents. Consequently, this

change will not create the possibility of a

new or different kind of accident from any

kind of accident previously evaluated.

(3) Involve a significant reduction in a

margin of safety:

The design basis of auxiliary electrical

systems is to supply the required ES loads of

one Unit and safe shutdown loads of the

other two units. The underground power

path is adequate to ensure operability of the

ES loads during the outage of the yellow bus.

The reliability of the emergency power

system is not significantly affected by a one

time extension of allowable outage time for

the overhead power path. Therefore, there

will be no significant reduction in any

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412 Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: July 10,

1995

Description of amendment request:

The proposed amendment would

modify the technical specifications to

minimize the potential for boron

deletion of the reactor coolant system

(RCS) during startup of an isolated loop.

The changes would permit RCS loop

isolation only during Modes 5 and 6.

RCS loop isolation valves would be

required open with power removed

from each isolation valve operator

during Modes 1, 2, 3, and 4. Primary

grade water would be isolated from the

RCS during Modes 4, 5, and 6, except

during planned boron dilution or

makeup activities.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed amendment would modify

the method used to prevent an inadvertent

boron dilution event during hot shutdown,

cold shutdown and during refueling. An

uncontrolled boron dilution transient cannot

occur during this mode of operation.

Inadvertent boron dilution is prevented by

administrative controls which isolate the

primary grade water system isolation valves

from the Chemical and Volume Control

System, except during planned boron

dilution or makeup activities. Thus

unborated water can not be injected into the

reactor coolant system, making an unplanned

boron dilution at these conditions highly

improbable, since the source of unborated

water to the charging pumps is isolated. This

precludes the primary means for an

inadvertent boron dilution event in this

mode of operation.

The primary grade water system isolation

valves may be opened when directed by the

control room during this mode of operation

only for a planned boron dilution or makeup

activity. The primary grade water system

isolation valves will be verified to be locked,

sealed or otherwise secured in the closed

position after the planned boron dilution or

makeup activity is completed. During

planned boron dilution events, operator

attention will be focused on the boron

dilution process and any inappropriate

blender operation will be readily identified.

The operator has prompt and definite

indication of any boron dilution from the

audible count rate instrumentation supplied

by the source range nuclear instrumentation.

High count rate is alarmed in the reactor

containment and the control room. In

addition a high source range flux level is

alarmed in the control room. The count rate

increase is proportional to the subcritical

multiplication factor.

The proposed amendment would also

modify the method used to prevent an

adverse reactor transient during startup of an

isolated reactor coolant loop. Procedures

require that the isolated loop water boron

concentration be verified prior to opening

loop isolation valves. Procedures also require

an isolated loop to be drained and refilled

from water supplied from the Refueling

Water Storage Tank (RWST) or Reactor

Coolant System (RCS) prior to opening either

the hot or cold leg isolation valves. Using

water from the RWST or RCS ensures 1) that

the boron concentration of the isolated loop

is sufficient to prevent a dilution of the active

reactor coolant loops and reducing the

shutdown margin to below those values used

in safety analyses when the isolated loop is

returned to service, and 2) that no single

failure could cause an isolated loop to be

filled with unborated water.

Thus procedures and interlocks prevent

inadvertent opening of loop isolation valves

and require that the startup of an isolated

loop be performed in a controlled manner

that virtually eliminates any sudden positive

reactivity addition from boron dilution. Thus

the core cannot be adversely affected by the

startup of an isolated loop and fuel design

limits are not exceeded. Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes do not create the

possibility of a new or different kind of

accident. No new systems, structures or

components are being proposed. Acceptable

alternative administrative controls are being

proposed to address inadvertent boron

dilution and the startup of inactive reactor

coolant loops.

The primary source of unborated water

will be isolated from injecting by the

charging pumps into the reactor coolant

system during hot shutdown, cold shutdown,

and refueling, except for planned boron

dilution events and makeup activities. The

proposed administrative controls prevent the

possible accident previously evaluated, i.e.,

an inadvertent boron dilution event.

A currently installed interlock to

recirculate reactor coolant in an isolated loop

is proposed to be deleted. In its place, each

reactor coolant isolated loop will be drained

and refilled with water supplied from the

RWST just before the loop is returned to

service. This administrative control will

prevent any inadvertent reactivity transient

when returning the loop to service. Thus, the

proposed administrative controls will

prevent the type of accident previously

evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed changes will continue to

ensure that adequate protection is provided

against an inadvertent boron dilution and the

adverse effects from the startup of an isolated

reactor coolant loop. General Design Criteria

10 requirements will not be exceeded with

respect to demonstrating specified acceptable

fuel design limits. The required indications

and functions are still maintained in

accordance with current technical

specification requirements and the shutdown

margin is unaffected. Therefore, the proposed

change will not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. Library, 663 Franklin

Avenue, Aliquippa, Pennsylvania

15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42603

Duquesne Light Company, et al., Docket

No. 50-334, Beaver Valley Power

Station, Unit 1, Shippingport,

Pennsylvania

Date of amendment request: July 11,

1995

Description of amendment request:

The proposed amendment would revise

the required area of the Reactor Coolant

System (RCS) overpressure protection

system vent from 3.14 square inches to

2.07 square inches. This vent is

provided to relieve a potential RCS

overpressure condition if the poweroperated relief valves (PORVs) are not

operable. The proposed vent area is

equal to the relief area of a PORV. A

single PORV is capable of providing

sufficient relief capacity to mitigate

potential low temperature

overpressurization events.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change is considered to be

editorial since it replaces the 3.14 square

inch vent size stated in overpressure

protection system (OPPS) Specifications

3.4.9.3, 3.1.2.1.b, and 3.1.2.3 and Bases 3/

4.1.2 and 3/4.4.9 with a 2.07 square inch vent

size. This ensures the vent size stated in the

technical specifications is consistent with the

actual size of an installed PORV. These

changes maintain consistency with the

analyses assumptions and the operation of

the OPPS in accordance with applicable

analyses and the UFSAR [Updated Final

Safety Analyses Report]. Therefore, we have

concluded that these changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated in the UFSAR.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes do not involve any

physical changes to the OPPS or their

setpoints. These changes do not change any

function previously provided by the OPPS.

These changes do not affect any failure

modes defined for any plant system or

component important to safety nor has any

new limiting single failure been identified as

a result of these changes. Therefore, these

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated in the UFSAR.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed changes will not affect the

operation of or the reliability of the OPPS.

These changes do not affect the manner in

which the plant is operated or involve a

change to equipment or features that affect

the operational characteristics of the plant.

Therefore, operation of the plant in

accordance with the proposed amendment

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: July 20,

1995

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.8.1.1 to

incorporate guidance provided in NRC

Generic Letter (GL) 84-15, ‘‘Proposed

Staff Actions to Improve and Maintain

Diesel Generator Reliability,’’ and GL

93-05, ‘‘Line-Item Technical

Specification Improvements To Reduce

Surveillance Requirements For Testing

During Power Operation,’’ which

includes (1) revised requirements for

testing the operable emergency diesel

generators (EDGs) for various

combinations of inoperable offsite

circuits and EDGs and (2) revised

surveillance requirements for the EDGs.

The revised surveillance requirements

include specifying generator voltage,

frequency limits, and diesel starting

time. In addition, several editorial

changes would be made to TS 3/4.8.1.1

which would be consistent with the

guidance provided in the NRC’s

Improved Standard Technical

Specifications (NUREG-1431).

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The probability of occurrence of a

previously evaluated accident is not

increased because the allowable outage times

for the offsite circuits and diesel generators

remain unchanged. The consequences of an

accident previously evaluated is not

increased because reducing the diesel

generator test frequency and permitting

additional test evolutions are intended to

minimize diesel wear and mechanical stress.

By eliminating excessive testing, which can

lead to premature diesel failures and

minimizing diesel wear and mechanical

stress, the diesel generator reliability is

increased. The consequences of an accident

previously evaluated is also not increased

because the addition of the parameters for

generator voltage, frequency, and diesel

starting time to the surveillance requirement

will provide additional assurance that the

diesel generators are performing as assumed

in the safety analysis. This proposed change

does not affect the availability or reliability

of the offsite circuits.

Therefore, this change will not increase the

probability or consequences of an accident

previously evaluated due to the continued

availability and reliability of the A.C.

electrical power sources.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes do not alter the

method of operating the plant. The changes

do not introduce any new failure modes and

are intended to increase the diesel generator

reliability and provide additional assurance

that the diesels are performing as assumed in

the safety analysis. The revision to the

various action statements and surveillance

requirements provide assurance that the

diesel generators will be able to power their

respective safety systems if required. The

proposed changes do not impact the

performance of any safety system.

Therefore, this proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The margin of safety is not reduced

because the A.C. electrical power sources

will continue to provide sufficient capacity,

capability, redundancy, and reliability to

ensure availability of necessary power to

engineered safety feature (ESF) systems. The

ESF systems will continue to function, as

assumed in the safety analyses, to ensure that

the fuel, reactor coolant system and

containment design limits are not exceeded.

The elimination of excessive testing on the

diesel generators are permitting additional

test evolutions, which result in less diesel

wear and mechanical stress, are intended to

increase diesel reliability. The increased

reliability of the diesels adds to the ability of

the A.C. electrical power source to provide

power to ESF systems. The proposed

additions to the surveillance requirements

will provide additional assurance of the

ability of the A.C. electrical power sources to

provide power to ESF systems.

Therefore, this proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the
42604 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket

No. 50-412, Beaver Valley

PowerStation, Unit 2, Shippingport,

Pennsylvania

Date of amendment request: July 24,

1995

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.4.11,

‘‘Relief Valves,’’ and associated Bases to

make Unit 2 TS 3/4.4.11 consistent with

Unit 1 TS 3/4.4.11, which was revised

by Unit 1 License Amendment No. 187

issued on May 15, 1995. The proposed

amendment would also generally reflect

the guidance provided in NRC Generic

Letter 90-06 and in the NRC’s Improved

Standard Technical Specifications

(NUREG-1431).

Basis for proposed no significant

hazards considerationdetermination: 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?

Implementation of these changes will

increase the availability of the poweroperated relief valves (PORVs) and their

associated block valves. The increased

availability is obtained through maintaining

power to the block valves which are closed

to control PORV seat leakage. Maintaining

power to the block valve provides the

flexibility of reopening the valves to control

reactor coolant system pressure. The

proposed change modifies Specification

3.4.11 actions, a surveillance requirement,

and Bases to generally reflect the

requirements of Generic Letter (GL) 90-06,

and the guidance provided in NUREG-1431,

‘‘Improved Standard Technical

Specifications’’ (ISTS) and is consistent with

the changes the NRC approved for Unit No.

1. A revised stress analysis has been

completed that takes credit for the speed at

which the block valve opens when manually

reducing reactor coolant system pressure.

The block valve relatively slow opening

speed reduces the peak pressure surge and

results in acceptable downstream piping

stress values. The PORV downstream piping

has been evaluated assuming manual vent

path operation with cold loop seal slug flow

and it has been determined that the piping

supports can accept these design transient

loads. The proposed change to the action

statement to close the block valve to isolate

a PORV and maintain power to the block

valve does not significantly increase the

probability of a small break loss of coolant

accident. No PORV function has been deleted

and the PORV and block valve continue to

be capable of being manually closed at any

time. As a result of the change to action ‘‘a,’’

an exception to the stroking requirements is

no longer required, therefore, reference to

action ‘‘a’’ in Surveillance Requirement

4.4.11.2 has been deleted. Closing the block

valve for a PORV that is not capable of being

manually cycled and removing power to the

block valve assures that the valve will not be

inadvertently opened when the condition of

the PORV is uncertain.

The changes remain consistent with the

analysis assumptions regarding the operation

of the PORVs and block valves and provides

increased assurance of their availability in

mitigating the consequences of a steam

generator tube rupture (SGTR) accident. The

requirements of GL 90-06 are substantially

addressed in the ISTS which have been

incorporated here except for specific design

differences. Minor editorial changes

involving capitalization have been

incorporated to maintain the format and

content and do not affect any of the

requirements, the accident analyses, or the

operation of the plant. Therefore, we have

concluded that these changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated in the UFSAR [Updated Final

Safety Analysis Report].

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes to the action

statements for the PORVs and the associated

block valves will improve the availability of

these valves for normal operation and for

mitigation of a SGTR accident. The proposed

changes do not involve any physical changes

to the PORVs or their setpoints. These

changes do not delete any design basis

accident function previously provided by the

PORV vent path nor has the probability of

inadvertent opening been increased.

Accordingly, no new limiting single failure

has been identified as a result of these

changes. Therefore, these changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated in the UFSAR.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed changes have been

incorporated to provide the capability to

manually stroke the vent path using the block

valve to control the pressure surge as a PORV

opens. The resultant downstream piping

forces were found acceptable, therefore,

power can be maintained to the block valve

when the block valve has been closed to

isolate a PORV because of excessive seat

leakage. This will allow operation of the

PORVs in a manner similar to the guidance

provided in GL 90-06 to improve PORV

availability. These changes will improve the

operator use of an isolated PORV since it is

now analyzed to be manually cycled with the

block valve closed and power maintained so

the operator can use the PORV if required to

mitigate the effects of a SGTR accident. This

is consistent with the intent of the ISTS and

does not affect the UFSAR, therefore,

operation of the plant in accordance with the

proposed amendment would not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 1500l.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: April 4,

1995

Description of amendment request:

The proposed amendment revises the

minimum water level that is required to

be maintained over irradiated fuel

assemblies during latching and

unlatching of control element

assemblies.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The fuel handling accident analysis

assumes that a fuel assembly is dropped

during fuel handling. During the latching and

unlatching of the CEAs, the upper guide

structure is in place and the CEDM extension

shaft assemblies are disconnected from their

CEA for subsequent removal with the vessel

upper guide structure. The dropping of a

CEA from the maximum height of six inches

will not damage that particular fuel assembly

or any surrounding fuel assemblies since this

movement is confined to within the upper

guide structure and the guide tubes of the

associated fuel assembly during this activity.

This less than six inches of movement does

not have the potential to result in a fuel

handling accident; therefore, an increase in

the probability of this accident does not

occur. The requirement to have at least 23

feet of water over the top of the irradiated

fuel assemblies during fuel and CEA

movement ensures that, should a fuel

handling accident occur, the resulting offsite

dose consequences are mitigated. The six

inch movement of the CEA during CEA

decoupling does not constitute fuel or CEA
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42605

movement which would result in a fuel

handling accident. As such, Technical

Specifications are unchanged with respect to

the mitigating requirements for a fuel

handling accident.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed change does not change the

design, configuration, or method of operation

of the plant; therefore, it does not create the

possibility of a new or different kind of

accident. Because no new equipment is being

introduced, and no equipment is being

operated in a manner inconsistent with its

design, the possibility of equipment

malfunction is not increased. The proposed

change adds an exception to the applicability

section and is bounded by the existing fuel

handling accident analysis.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

There is no reduction in margin of safety

in that 23 feet of water is still maintained

over the irradiated fuel assemblies anytime

there is a potential for a fuel handling

accident. Adding the exception of the

latching and unlatching of the CEAs to the

applicability section does not involve a

change in the accident analysis for fuel

handling which remains bounding.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: July 21,

1995

Description of amendment request:

The proposed change requests that the

current expiration date for license NPF29 be changed to reflect the issuance

date of the new license granted Grand

Gulf on November 1, 1984. The change

consists of extending the expiration date

to 40 years from the date of issuance of

license NPF-29 (November 1, 1984 to

November 1, 2024).

Basis for proposed no significant

hazards considerationdetermination: 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:

a. No significant increase in the probability

or consequences of an accident previously

evaluated results from this change.

The proposed change does not affect the

design or operation of any plant system. The

effect of 40 years of full power operations has

previously been evaluated and documented

in the Updated Final Safety Analysis Report

(UFSAR). The design life of structures,

systems and components is controlled by

existing plant problems [sic., programs] and

processes that are not affected by this change.

The proposed change will simply allow

Grand Gulf to achieve its original planned 40

years of service. Equipment associated with

initiating event frequencies or accident

mitigation must continue to meet all

applicable maintenance and operability

requirements regardless of license duration

(It is also interesting to note that the license

duration limitation of 40 years, as contained

in 10 CFR 50.51 is not a limitation resulting

from concerns over plant aging effects. ‘‘In

fact, the limit was a compromise between the

efforts of the Justice Department and electric

cooperatives, who championed a 20-year

limit on the basis of antitrust concerns, and

the view of the utility industries that a longer

period was necessary to ensure full

amortization of a nuclear power plant.’’ (56

FR 64961, December 13, 1991)). Therefore,

the probability or consequences of previously

analyzed accidents are not significantly

increased.

b. The change would not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed change will not add any

plant equipment or introduce any new modes

of plant operation. The change will only

amend the operating license to allow 40 years

of full power operations. The proposed

change does not affect the current

maintenance or surveillance practices, which

are designed to maintain and monitor the

current service life of plant structures,

systems and components in accordance with

regulatory requirements. Therefore, the

proposed change does not create the

possibility of new equipment failure modes

or a new or different kind of accident from

any accident previously evaluated.

c. The change would not involve a

significant reduction in a margin of safety.

The proposed change does not involve a

significant reduction in a margin of safety

since it only provides for 40 years of full

power operations for which the plant is

designed. Current Technical Specification

surveillance requirements (e.g. associated

with 10 CFR 50 Appendix H) and other

regulatory requirements remain in place and

will ensure continued compliance with

applicable safety margins.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: June 20,

1995

Description of amendment request:

The proposed Technical Specifications

(TS) changes would remove the

surveillance interval text for the 10 CFR

Part 50, Appendix J, Type A test

(Integrated Leak Rate Test or ILRT), and

Drywell-to-Suppression Chamber

(bypass) leakage test specified in TS

Surveillance Requirements (SR)

4.6.1.2.a, 4.6.1.2.b, and 4.6.2.1.e.

Basis for proposed no significant

hazards considerationdetermination: 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 TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The primary containment and the

suppression chamber are not considered to

be accident initiators, they are accident

mitigators. There are no physical or

operational changes to the containment or

suppression structure, system or components

being made as a result of the proposed

changes. These changes will not impose

different requirements and adequate control

of information will be maintained. These TS

changes will not alter assumptions made in

the safety analysis and licensing basis.

Therefore, the proposed TS changes to

eliminate the details of the test intervals will

not increase the probability or consequences

of an accident previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes remove the specific

surveillance test interval text from TS and

address the interval by direct reference to the

applicable regulation. The proposed TS

changes do not make any physical or

operational changes to existing plant systems

or components. Furthermore, the primary

containment and suppression chamber act as
42606 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

accident mitigators not initiators. Therefore,

the possibility of a new or different kind of

accident than from any accident previously

evaluated is not introduced.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

LGS [Limerick Generating Station] TS

Bases 3/4 6.1.2 state that surveillance testing

is consistent with 10 CFR 50, Appendix J and

does not specify a SR test interval. TS Bases

3/4 6.2, describing the bypass test does not

specify a SR test interval. However, the NRC

Safety Evaluation related to amendment Nos.

68 (Unit 1) and 31 (Unit 2) concluded that

it is acceptable for the drywell-tosuppression chamber test frequency to

coincide with the 10 CFR 50, Appendix J,

Type A test, since individual vacuum breaker

leakage tests are an acceptable alternative to

an integrated suppression pool bypass test

during outages for which a Type A

containment integrated leak rate test is not

conducted. The alternative bypass test

requirement, TS SR 4.6.2.1.f, is not affected

by these changes.

The Type A test, and bypass SR test

intervals are adequately presented in the test

implementing procedures, and TS will

directly reference 10 CFR 50, Appendix J, for

the appropriate test interval.

Therefore, the proposed TS 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

Sr., Esquire, Sr. V. P. and General

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Power Authority of The State of New

York, Docket No. 50-286, Indian

PointNuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: July 21,

1995

Description of amendment request:

The proposed amendment would

change Technical Specifications Section

6.0 (Administrative Controls) to replace

the title-specific list of members on the

Plant Operating Review Committee

(PORC) with a more general statement of

membership requirements. The scope of

disciplines represented on the PORC

would also be expanded to include

nuclear licensing and quality assurance.

The proposed amendment would also

change the title ‘‘Resident Manager’’ to

‘‘Site Executive Officer.’’ This title

change would not affect the reporting

relationship, authority, or responsibility

of the position.

Basis for proposed no significant

hazards considerationdetermination: 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:

Operation of the Indian Point 3 Nuclear

Power Plant in accordance with the proposed

amendment would not involve a significant

hazards consideration as defined in 10 CFR

50.92, since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes are administrative

in nature and do not involve plant equipment

or operating parameters. There is no change

to any accident analysis assumptions or other

conditions which could affect previously

evaluated accidents. The proposed changes

will not decrease the organization’s ability to

respond to a design basis accident.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated.

Since the proposed changes are

administrative in nature and do not involve

hardware design, modifications or operation,

the possibility of new or different accidents

is not created.

3. Involve a significant reduction in the

margin of safety.

The proposed title change for the Resident

Manager is an administrative change and

does not affect the responsibilities, authority,

or reporting relationships for this

management position. Replacing the title

specific list of PORC members with a

statement of membership requirements for

the committee does not reduce the

effectiveness of the committee to advise the

Resident Manager (Site Executive Officer) on

matters regarding nuclear 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: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: Ledyard B.

Marsh

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: March

30, 1995

Description of amendment request:

The proposed change to the Technical

Specifications (TS) would change TS

Table 3.3.1-2, ‘‘Reactor Protection

System Response Times’’, TS Table

3.3.2-3, ‘‘Isolation System

Instrumentation Response Time’’, TS

Table 3.3.3-3, ‘‘Emergency Core Cooling

System Response Times’’, and

associated Bases. The proposed changes

to the above-referenced TS Tables

would eliminate the requirement to

perform response time testing for certain

classes of equipment.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The purpose of the proposed Technical

Specification change is to eliminate response

time testing requirements for selected

instrumentation in the Reactor Protection

System, Isolation System, and Emergency

Core Cooling System. However, because of

the continued application of other existing

Technical Specification requirements such as

channel calibrations, channel checks,

channel functional tests, and logic system

functional tests, the response time of these

systems will be maintained within the

acceptance limits assumed in plant safety

analyses and required for successful

mitigation of an initiating event. The

proposed Technical Specification changes do

not affect the capability of the associated

systems to perform their intended function

within their required response time.

The BWR Owners’ Group has completed an

evaluation (NEDO-32291, ‘‘System Analyses

for the Elimination of Selected Response

Time Testing Requirements’’) which

demonstrates that response time testing is

redundant to the other Technical

Specification requirements listed in the

preceding paragraph. These other tests are

sufficient to identify failure modes or

degradation in instruments response time

and ensure operation of the associated

systems within acceptance limits. There are

no known failure modes that can be detected

by response time testing that cannot be

detected by the other Technical Specification

tests. Hope Creek Generating Station is

specifically bounded by the assumptions and

justifications in General Electric Company

Licensing Topical Report, NEDO-32291,

‘‘System Analyses for Elimination of Selected

Response Time Testing Requirements.’’

2. Will not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

As discussed above, the proposed

Technical Specification changes do not affect

the capability of the associated systems to

perform their intended function within the

acceptance limits assumed in plant safety

analyses and required for successful

mitigation of an initiating event. The

proposed elimination of response time

testing would not result in any new
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42607

equipment, operating modes, or plant

configurations.

3. Will not involve a significant reduction

in a margin of safety.

The current Technical Specification

response times are based on the maximum

allowable values assumed in the plant safety

analyses. These analyses conservatively

establish the margin of safety. As described

above, the proposed Technical Specification

changes do not affect the capability of the

associated systems to perform their intended

functions within the allowed response time

used as the basis for the plant safety analyses.

Plant and system response to an initiating

event will remain in compliance within the

assumptions of the safety analyses, and

therefore the margin of safety is not affected.

Although not explicitly evaluated, the

proposed Technical Specification changes

will provide an improvement to plant safety

and operation by:

a) Reducing the time safety systems are

unavailable

b) Reducing safety system actuations

c) Reducing shutdown risk

d) Limiting radiation exposure to plant

personnel

e) Eliminating the diversion of key

personnel to conduct unnecessary testing.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: April 18,

1995

Description of amendment request:

The proposed changes to the Technical

Specifications (TS) would change TS

Table 4.3.7.1-1 ‘‘Radiation Monitoring

Instrumentation Surveillance

Requirements.’’ This change would

increase the channel functional test

interval from monthly to quarterly for

each instrument.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change involves no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. Increasing

the interval between channel functional tests

for the radiation monitoring instrumentation

represent changes that do not affect plant

safety and do not alter existing accident

analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed change is procedural in

nature concerning the channel functional test

frequency for the radiation monitoring

instrumentation not already on a quarterly

surveillance. The channel functional test

methodology for these instruments remains

unchanged. The proposed changes, while

slightly increasing the possibility of an

undetected instrument error, will not create

a new or unevaluated accident or operating

condition.

3. Will not involve a significant reduction

in a margin of safety.

The proposed change is in accordance with

recommendations provided by the NRC

regarding the improvement of Technical

Specifications. These changes will result in

perpetuation of current safety margins while

reducing regulatory burden and decreasing

equipment degradation.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: May 4,

1995

Description of amendment request:

The proposed change to the Technical

Specifications (TS) would change TS 3/

4.6.1.8, ‘‘Drywell and Suppression

Chamber Purge System’’, to increase the

annual operational limit for the drywell

and suppression chamber purge system

from 120 to 500 hours.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change involves no

hardware changes and no changes to existing

structures. Increasing the annual operational

limit of the drywell and suppression chamber

purge system will not increase the

probability of a loss-of-coolant accident.

While increased usage of the purge system

will result in a slight increase in the

possibility that these valves will be open

during a LOCA, it will not alter or impact

previous LOCA analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed change will not result in an

unanalyzed condition. While the increase in

purge system operation will slightly increase

the possibility of the containment vent and

purge valves being open at the onset of a

LOCA event, the valves have been

established as capable of isolating the

containment within five seconds. This is well

within the bounds of existing LOCA analyses

which assume an open duration of 175

seconds. Therefore, this change will not

require a new or different accident analysis.

3. Will not involve a significant reduction

in a margin of safety.

The proposed change will not alter existing

systems, equipment, components, or

structures. The method of operating the

drywell and suppression chamber purge

system will not be altered by the increased

annual usage. While there is a slight increase

in the possibility of purge operations at the

onset of a LOCA, any resulting release would

be insignificant and bounded by existing

LOCA analyses. Operation of the drywell and

suppression chamber purge system based on

these proposed changes will remain within

the guidance provided in the NRC’s Branch

Technical Position CSB 6-4.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Saxton Nuclear Experimental

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

Saxton Nuclear Experimental Facility

(SNEF), Bedford County, Pennsylvania

Date of amendment request: June 2,

1995, as supplemented on June 23,

1995.

Description of amendment request:

The proposed changes to the technical

specifications are administrative in
42608 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

nature. The proposed amendment

would revise the organization structure

associated with the SNEF to allow

General Public Utilities Nuclear

Corporation resources to be applied to

SNEC activities within their normal

organizational structure; eliminating the

need to identify and compartmentalize

a portion of the organization as specific

to SNEC. The proposed amendment

would also revise the description and

drawing of the SNEF site to reflect

multiple gates in the SNEF fence.

Basis for proposed no significant

hazards considerationDetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below: The proposed changes do not

involve a significant hazards

considerations because the changes

would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The administrative changes will not impact

the physical condition of the containment

vessel as it relates to the risk of fire, flood or

radiological hazard.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

In its present condition, the only accidents

applicable to the site are those addressed

above.

3. Involve a significant reduction in a

margin of safety.

The proposed administrative changes

would have no effect on any margins of

safety for any evaluated accidents.

The NRC staff has reviewed the

analysis of the licensee and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

Location: Saxton Community Library,

911 Church Street, Saxton,

Pennsylvania 16678Attorney for the

Licensee: Ernest L. Blake, Jr., Esquire,

Shaw, Pittman, Potts, and Trowbridge,

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

20037

NRC Project Director: Seymour H.

Weiss

South Carolina Electric & Gas

Company, South Carolina Public

Service Authority, Docket No. 50-395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of amendment request: June 30,

1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) for the

pressurizer power operated relief valves

(PORVs) to follow the guidance of

Generic Letter (GL) 90-06, Generic Issue

70, and the improved Westinghouse

Standardized Technical Specifications

(NUREG-1431, Rev. 1).

Basis for proposed no significant

hazards considerationdetermination: 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 probability or consequences of an

accident previously evaluated is not

significantly increased.

There is no increase in the probability of

an accident because the physical

characteristics of the PORVs and their block

valves remain unchanged. No changes to any

hardware or software that affects these

components is planned.

The PORVs are pressure relieving devices

and only two failure modes need to be

considered. The first is that one or more

PORVs or block valves fail to open when

required. This is not

a significant concern and is not a credible

cause of any accident. The second mode is

failing to close which includes

depressurization of the RCS [reactor coolant

system] and a reactor trip on low pressurizer

pressure or overtemperature [delta]T. The

consequences for the more limiting

Pressurizer Safety Valve Accidental

Depressurization event has been analyzed

with acceptable results.

There is no increase in the consequences

of an accident as a result of this change,

because only one PORV is required to

mitigate the consequences of a design basis

Steam Generator Tube Rupture. There is

sufficient redundancy to ensure one PORV is

available to perform this function even if one

PORV is inoperable or incapable of being

manually cycled. The validation of the

Emergency Operating Procedures on the

VCSNS [Virgil C. Summer Nuclear Station]

simulator demonstrated that one pressurizer

PORV has sufficient capacity to depressurize

the RCS in a time frame which will not cause

the offsite doses presented in the FSAR

[Final Safety Analysis Report] to be

exceeded.

The PORVs are utilized to depressurize the

RCS and equalize the pressure between the

primary and secondary systems. This stops

the intrusion of RCS water into the secondary

which can be released into the atmosphere.

By the time the PORVs are called upon, the

affected steam generator (SG) has been

identified and steps have been taken to

isolate the faulted SG. This acts to minimize

the radiological impact on the health and

safety of the public. In all cases, the dose

results are within 10 CFR 100 limits.

2. The possibility of an accident or a

malfunction of a different type than any

previously evaluated is not created.

The proposed TSCR [TS Change Request]

does not involve any physical changes to the

plant or decrease the number of PORVs and

block valves that must be capable of

performing their intended function. These

components are used to mitigate the effects

of postulated events and their failure has

already been considered. The worst case

failure, either not opening or not closing, has

been evaluated and is bounded by other more

limiting accidents.

3. The margin of safety has not been

significantly reduced.

The currently approved TS permits all

three PORVs and/or their block valves to be

inoperable as long as precautions are taken

to assure that RCS would not leak-by,

assuming single failures and spurious

operation. The proposed TSCR would require

a minimum of two PORVs and block valves

to be operable, or at least capable of being

manually cycled, in Modes 1, 2, and 3. This

is in fact an increase in margin and provides

for greater reliability with the added benefit

that the probability of challenges to the

pressurizer code safety valves will be

lessened.

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: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Project Director: Frederick J.

Hebdon

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

Authority, Docket No. 50-395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: July 28,

1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

exclude the requirement to perform the

slave relay test of the 36-inch

containment purge supply and exhaust

valves on a quarterly basis while the

plant is in Modes 1, 2, 3, or 4.

Basis for proposed no significant

hazards considerationdetermination: 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?

No, the probability or consequences of an

accident previously evaluated would not be

increased since no credit is taken for the

valves in FSAR [Final Safety Analysis

Report] Chapter 15.

The only credible accident discussed in

FSAR Chapter 15 that applies to these valves

is a fuel handling accident inside
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42609

containment (15.4.5.1). The analysis assumes

the escaped gases are released

instantaneously to the environment via the

Reactor

Building purge system. The analysis does

not take credit for these valves nor for

filtration or holdup time during release. The

result of the analysis is acceptable and offsite

doses are within the limits of 10 CFR 100.

TS 3.6.1.7 requires that these valves be

sealed shut during Modes 1, 2, 3, and 4.

When sealed shut, these valves will not open

via any signal.

With these valves already in a shut

position, neither the probability nor the

consequences of an accident are increased.

2. Does the change create the possibility of

a new or different kind of accident from any

previously evaluated?

No, the 36’’ [inch] containment purge

exhaust and supply valves will not be placed

in a condition different from that evaluated

previously.

The only credible accident discussed in

FSAR Chapter 15 that applies to these valves

is a fuel handling accident inside

containment (15.4.5.1). The analysis assumes

the escaped gases are released

instantaneously to the environment via the

Reactor Building purge system. The analysis

does not take credit for these valves nor for

filtration or holdup time during release. The

result of the analysis is acceptable and offsite

doses are within the limits of 10 CFR 100.

Additionally, TS 3.6.1.7. requires that

these valves be sealed shut during Modes 1,

2, 3, and 4. When sealed shut, these valves

will not open via any signal.

3. Does the change involve a significant

reduction in the margin of safety?

TS 4.3.2.1. requires that this slave relay test

be performed quarterly. This surveillance is

accomplished for the 36’’ [inch] containment

purge exhaust and supply valves by cycling

the respective K615 relay. This will not

provide assurance that the valve will perform

its safety function since the valve is sealed

closed. The proposed change will exclude

the requirement to perform the K615 relay

test (auto actuation logic and actuation relays

- slave relay test) on a quarterly basis while

the plant is in Modes 1, 2, 3,or 4.

TS 3.6.1.7. requires that these valves be

sealed shut during Modes 1, 2, 3, and 4.

When sealed shut, these valves will not open

via any signal. Since this relay would not be

needed to supply a signal to place these

valves in the closed position, the margin of

safety is not affected.

Based on the preceding analysis, SCE&G

has determined that this change does no

involve a significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260 and 50-296, Browns

Ferry Nuclear Plant, Units 1, 2 and 3,

Limestone County, Alabama

Date of amendment request: June 2,

1995 (TS 353)

Description of amendment request:

The proposed amendment supports

replacement of the existing power range

neutron monitoring equipment and

implements ARTS/MELLL [average

power range monitor and rod block

monitor technical specifications/

maximum extended load line limit]

analysis improvements.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Group A Changes: This proposed TS

change is associated with the NUMAC PRNM

[nuclear measurement analysis and control

power range neutron monitor] retrofit design.

The proposed TS change involves

modification of the LCOs [limiting condition

for operations] and SRs [surveillance

requirements] for equipment designed to

mitigate events which result in power

increase transients. For the APRM [average

power range monitor] system mitigative

action is to block control rod withdrawal or

initiate a reactor scram which terminates the

power increase when setpoints are exceeded.

For the RBM [rod-block monitor] system

mitigative action is to block continuous

control rod withdrawal prior to exceeding the

MCPR [minimum critical power ratio] safety

limit during a postulated Rod Withdrawal

Error [RWE]. The worst case failure of either

the APRM or the RBM systems is failure to

initiate mitigative action (failure to scram or

block rod withdrawal). Failure to initiate

mitigative action will not increase the

probability of an accident. Thus, the

proposed change does not increase the

probability of an accident previously

evaluated.

For the APRM and the RBM systems, the

NUMAC PRNM design, together with revised

operability requirements (LCOs) and revised

testing requirements (SRs), results in

equipment which continues to perform the

same mitigation functions under identical

conditions with reliability equal to or greater

than the equipment which it replaces.

Because there is no change in mitigation

functions and because reliability of the

functions is maintained, the proposed change

does not involve an increase in the

consequences of an accident previously

evaluated.

Group B Changes: This proposed change is

associated with implementation of the ARTS/

MELLL analysis. The proposed change will

permit expansion of the current allowable

power/flow operating region and will apply

a new methodology for assuring that fuel

thermal and mechanical design limits are

satisfied. Reference 3 evaluates operation in

the MELLL region with assumed

implementation of the ARTS changes. The

conclusion of reference 3 is that for all events

and parameters considered there is adequate

design margin for operation in the MELLL

region. Because operation in the MELLL

region maintains adequate design margin, the

proposed change does not significantly

increase the probability of an accident

previously evaluated.

In support of operation in the MELLL

region, the proposed change modifies flowbiased APRM scram and rod block setpoints

and implements new RBM power-biased

setpoints. This potentially changes the way

in which the APRM and RBM systems

perform their mitigation functions. However,

no credit for the flow-biased APRM scram or

rod block is taken in mitigation of any design

basis event; thus, changing the APRM

setpoints does not impact the consequences

of any accident previously evaluated. The

proposed changes to the RBM system

potentially impact mitigation of the RWE.

However, per discussion in reference 3, the

proposed RBM changes will assure that the

RWE is not a limiting event; thus, the

consequences of the RWE are not increased.

The proposed change does not increase the

consequences of an accident previously

evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes (Group A and Group

B) involve modification and replacement of

the existing power range neutron monitoring

equipment, modification of the setpoints and

operational requirements for the APRM and

RBM systems, implementation of a new

methodology for administering compliance

with fuel thermal limits, and operation in an

extended power/flow domain. These

proposed changes do not modify the basic

functional requirements of the affected

equipment, create any new system interfaces

or interactions, nor create any new system

failure modes or sequence of events that

could lead to an accident. The worst case

failure of the affected equipment is failure to

perform a mitigation action, and failure of

this mitigative equipment does not create the

possibility of a new or different kind of

accident. The proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

Group A Changes: This proposed TS

change is associated with the NUMAC PRNM

retrofit design. The NUMAC PRNM change

does not impact reactor operating parameters

nor the functional requirements of the power
42610 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

range neutron monitoring system. The

replacement equipment continues to provide

information, enforce control rod blocks and

initiate reactor scrams under appropriate

specified conditions. The proposed change

does not revise any safety margin

requirements. The replacement APRM/RBM

equipment has improved channel trip

accuracy compared to the current system and

meets or exceeds system requirements

previously assumed in setpoint analysis.

Thus, the ability of the new equipment to

enforce compliance with margins of safety

equals or exceeds the ability of the

equipment which it replaces. The proposed

change does not involve a reduction in a

margin of safety.

Group B Changes: This proposed change is

associated with implementation of

recommendations presented in the ARTS/

MELLL analysis. Operation in the MELLL

region does not affect the ability of the plant

safety-related trips or equipment to perform

their functions, nor does it cause any

significant increase in offsite radiation doses

resulting from any analyzed event. Analyses

documented in reference 3 demonstrate that

for operation in the MELLL region adequate

margin to design limits is maintained.

Implementation of the ARTS improvements

provides flow- and power-dependent thermal

limits which maintain existing margins of

safety in normal operation, anticipated

operational occurrences and accident events.

Implementation of power-biased RBM

setpoints improves the margin of safety in a

postulated RWE by assuring that the RWE is

not a limiting event. The proposed change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260 and 50-296, Browns

Ferry Nuclear Plant, Units 1, 2 and 3,

Limestone County, Alabama

Date of amendment request: June 8,

1995 (TS 361)

Description of amendment request:

The proposed amendment clarifies the

definition of operability for the RHRSW

system standby coolant supply

capability and revises the instrument

numbers for several instruments that

have been upgraded.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to TS 3.5.C.3 clarifies

the operability requirements of the standby

coolant supply capability. It does not change

or degrade the nuclear safety characteristics

of the RHRSW and RHR systems and will not

affect the intent of the TS. The operation of

the standby coolant supply capability is not

a precursor to any design basis accident or

transient analyzed in the BFN FSAR. The

proposed changes to instrument numbers are

administrative changes for the upgraded

drywell temperature and pressure

instrumentation. The proposed changes do

not affect the design basis or the safety

functions of the Primary Containment

system, since the function and

instrumentation range is not changed.

Therefore, the probability of occurrence or

the consequences of an accident or

malfunction of equipment important to safety

previously evaluated in the safety analysis

report has not been increased.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The possibility for an accident or

malfunction of a different type than any

evaluated previously in the safety analysis

report is not created by this change. The

change to TS 3.5.C.3 adds the indication of

associated valves of the function involved

and a clarification of operability for the

standby coolant supply connection to be

commensurate with the RHR cross-connect

capability. The proposed changes to

instrument numbers are administrative

changes effected by the upgrade of

instrumentation. There are no automatic

actions affected or compromised by these

changes.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change to TS 3.5.C.3 does

not affect any acceptable limit of operation or

analysis assumption in the TS or Bases. The

changes affect neither setpoints, calibration

intervals, nor functional test intervals. The

change does not affect any acceptable limit

of operation or analysis assumption found in

the TS or their bases. The proposed

administrative changes to the instrument

numbers do not affect the setpoint,

calibration interval or function of the

instrumentation. These changes do not affect

any limiting conditions of operation or

analysis assumption in the TSs or their bases.

Therefore, the change does not reduce the

margin of safety as defined in the basis for

any TS.

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: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260 and 50-296, Browns

Ferry Nuclear Plant, Units 1, 2 and 3,

Limestone County, Alabama

Date of amendment request: June 16,

1995 (TS 360)

Description of amendment request:

The proposed change will revise the

BFN Units 1, 2, and 3 Technical

Specifications (TS) to permit the

Traversing In-Core Probe (TIP) system to

be considered operable with less than

five TIP machines operable. The

proposed amendment will allow the

utilization of substitute data in lieu of

data from inaccessible TIP measurement

locations. The substitute data will be

derived from either symmetric TIP

measurement locations (under certain

core conditions) or from normalized TIP

data as calculated by the on-line core

monitoring system.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The TIP system is not used to prevent, or

mitigate the consequences of any previously

analyzed accident or transient; nor are any

assumptions made in any accident analysis

relative to the operation of the TIP system.

The primary containment isolation function

(TIP withdrawal) is not affected. The

proposed TS change does not alter the

fundamental process involved in calibrating

neutron instrumentation (LPRMs) [local

power range monitors], but requires that only

the equipment associated with the TIP

channels necessary for recalibrating LPRMs

and for core monitoring functions be

operable. Collection and storage of TIP data

without using all TIP channels is acceptable

because TIP machine normalization factors

are ultimately derived from the most recent

full core TIP set, which intercalibrates the

TIP machines in a common core location.

Additionally, the use of symmetric

detectors and analytical values as substitute

data for inaccessible TIP channels does not

compromise the ability of the process

computer to accurately represent the spatial

neutron flux distribution of the reactor core.
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42611

The core monitoring methodology is

presently based on symmetry of rod patterns

and fuel loading. This is not changed but

extended to use a higher order of symmetry

(octant symmetry) which exists with ‘‘type

A’’ sequence rod patterns. Therefore, this

change does not increase the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve the

installation of any new equipment, or the

modification of any equipment designed to

prevent or mitigate the consequences of

accidents or transients. Therefore, the

proposed amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The total core TIP reading uncertainties

will remain within the assumptions of the

licensing basis. Therefore, the margin of

safety to the MCPR [minimum critical power

ratio] safety limits is not reduced. The ability

of the process computer to accurately

represent the spatial neutron flux

distribution for the reactor core is not

compromised. Additionally, the computer’s

ability to accurately predict the LHGR [linear

heat generation rate], APLHGR [average

planar linear heat generation rate], MCPR and

its ability to provide for LPRM calibration is

not compromised. 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: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50-440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: October

21, 1994

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.6.1.2,

‘‘Primary Containment Leakage.’’ The

changes would clarify that the main

steam line isolation valves leakage is

accounted for separately from the

integrated primary containment leak

rate or combined local leak rate results.

Also, two references would be deleted,

the test duration for use of Bechtel

Corporation Topical Report BN-TOP-1

would be clarified, and the requirement

to perform the third integrated leak rate

in each 10-year service period in

conjunction with the 10-year plant

inservice inspection would be deleted.

Exemptions to 10 CFR Part 50 Appendix

J, ‘‘Primary Reactor Containment

Leakage Testing for Water-Cooled Power

Reactors,’’ are also being requested in

conjunction with the proposed TS

changes.

Basis for proposed no significant

hazards considerationdetermination: 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:

Part A - Formalize the Approval for

Excluding the Main Steam Line Isolation

Valve Leakages from Inclusion in i) the

Overall Integrated Primary Containment Leak

Rate and ii) the Combined Local Leak Rate,

and Clarify that the Main Steam Lines are

Not Required to be Vented and Drained for

Type A Testing

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Since Appendix J was originally

envisioned, alternative means of meeting the

intent of these requirements have been

developed which provide an equivalent level

of protection of the public health and safety.

However, since some of these alternatives

deviate from the specific wording of

Appendix J, exemptions are appropriate for

these alternatives. Implicit in the FSAR

treatment of the main steam line leakage, as

well as the TS requirements for main steam

line leakage, are several deviations from the

specific requirements of Appendix J.

Although PNPP’s methods and practices for

Appendix J testing have been previously

described in correspondence to the NRC, a

formal exemption was not recognized to be

needed at that time in that the NRC’s

approval was perceived to be received by the

issuance of the PNPP TS. Exemption to four

separate paragraphs of 10 CFR 50 Appendix

J will document the approvals previously

received and incorporated into the TS for

main steam line isolation valve testing during

the initial licensing of the PNPP. This TS

change adds references to footnotes within

the TS LCO 3.6.3.1 to clarify which

conditions represent exemptions to

Appendix J. These exemptions are described

in the Bases.

PNPP utilized the criteria described in the

Standard Review Plan (SRP), Section 15.6.5,

Appendix D, ‘‘Radiological Consequences of

a Design Basis Loss-of-Coolant Accident:

Leakage from Main Steam Isolation Valve

Leakage Control System (Rev. 1 - July 1981).’’

This is an alternative, NRC approved method

for assessing the MSIV leakage contribution

and determining the radiological

consequences.

In accordance with the SRP, the safety

analysis for a design basis LOCA includes the

maximum main steam line leak rate

separately from the maximum containment

leak rate. Within Appendix J it is implied

that Type A tests are intended to measure the

primary containment overall integrated leak

rate, but this vas before the SRP Section was

developed which allows the MSIV

contribution to be accounted for separately in

the safety analysis. Therefore, the MSIV leak

rate should not be included in the

measurement of the ILRT. Including the

MSIV leakage in the combined local leak rate

limit is also not necessary since a specific

Type C MSIV leak rate has been specified in

TS 3.6.1.2.

In summary, there is no change in the

probability or consequences of any accident

since the addition of the references and

footnotes to clarify the TS LCO and Actions

do not change the design of the plant, nor the

operational characteristics of any plant

system, nor the procedures by which the

Operators run the plant. These changes only

cite formal Appendix J exemptions which are

requested to document the approval

previously received. A formal request for

exemption to the applicable paragraphs of 10

CFR 50 Appendix J is also being submitted

in a separate letter in conjunction with this

proposed TS change.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. There are no design changes being

made that would create a new type of

accident or malfunction, and the method and

manner of plant operation remains

unchanged. The only change being made is

an exemption to 10 CFR 50 Appendix J

which will be cited in the TS to document

the implicit and explicit approvals of the

PNPP design and testing methods for main

steam line isolation valves. The requirements

and bases for which the formal exemption is

sought are currently presented and

implemented in the licensing basis and the

TS for PNPP. The objective of the regulation

is being met and will continue to be met. The

exemption to 10 CFR 50 Appendix J is being

submitted in a separate letter in conjunction

with this proposed TS change.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

These changes do not involve a significant

reduction in the margin of safety because

they are administrative in nature. The

proposed change will only cite the NRC

exemption that grants the deviation from

Appendix J. The proposed changes do not

affect any USAR design bases or accident

assumptions. Therefore, the proposed

changes do not reduce the margin of safety

as defined in the bases for any Technical

Specification.

Part B - Revise Surveillance Requirement

4.6.1.2 to Eliminate Unnecessary References

and ClarifY the Use of BN-TOP-1
42612 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Surveillance Requirement 4.6.1.2 is

proposed to be revised to eliminate the direct

reference to the ANSI Standards N45.4 and

N56.8 within the text, because these same

Standards are listed within Appendix J. It is

unnecessary to repeat the references to the

Standards within the Technical

Specifications because the PNPP is still

required to be in compliance with the

regulations. No additional benefits are gained

and licensee flexibility to upgrade to later

versions of the Standards is reduced since a

Technical Specification change is necessary

to change the version of the Standard to

which PNPP is committed. This change

removes a redundant requirement to list

these Standards in the Technical

Specifications. Therefore, this change cannot

involve a significant increase in the

probability or consequences of an accident

because the regulation is still required to be

met.

A reference to Topical Report BN-TOP-1

continues to be retained within Surveillance

Requirement 4.6.1.2, and the use of the report

is clarified to be for test durations less than

24 hours. This reference is retained within

the TS since a reference to BN-TOP-1, though

not specifically included within Appendix J,

is allowed by Section 7.6 of ANSI N45.4-1972

and has been approved for PNPP use by the

NRC. The TS Bases are also proposed to be

revised to include a statement that the use of

BN-TOP-1 is in accordance with Appendix J.

These changes result in no changes to plant

systems and have no effect on accident

conditions or assumptions. These proposed

changes do not affect possible initiating

events for accidents previously evaluated, or

any system functional requirements. Hence,

these changes are purely administrative in

that they are designed to eliminate a

redundant requirement and clarify the

applicability and acceptability of an

alternative leak rate testing provision within

the TS. These changes do not affect plant

operation in any way. Therefore, the

proposed changes do not affect the

probability or consequences of any 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.

There are no design changes being made

that would create a new type of accident or

malfunction, and the method and manner of

plant operation remains unchanged. These

changes eliminate a redundant requirement

and clarify the applicability and acceptability

of alternative leak rate testing provisions

within the TS. Since the alternative leak rate

testing provisions have been approved by the

NRC, the objective of the regulation

continues to be met. Therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

These changes do not involve a significant

reduction in the margin of safety because

they are administrative in nature and either

eliminate a redundant requirement or clarify

the applicability and acceptability of an

alternative, NRC approved, leak rate testing

provision within the TS. The proposed

changes do not affect any USAR design bases

or accident assumptions. Therefore, the

proposed changes do not reduce the margin

of safety as defined in the Bases for any

Technical Specification.

Part C - Decouple Performance of the Third

Type A Test from the Shutdown for the 10-

Year Plant Inservice Inspection

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change revises Surveillance

Requirement 4.6.1.2.a by removing the

second sentence requiring that the third test

of each containment Integrated Leak Rate

Test (ILRT) set be conducted during the

shutdown for the 10-year plant inservice

inspection. A request for an exemption to 10

CFR 50 Appendix J, Paragraph III.D.l(a) is

also being submitted in conjunction with this

proposed change. Note that this change is

also included in the proposed Appendix J

rule changes currently under consideration

and has been approved for several other

plants. The deletion of this requirement from

the Technical Specifications does not impact

plant safety because the 10 CFR 50 Appendix

J requirement that three Type A containment

ILRT tests to be performed over a 10 year

period is not affected. This change only

removes an unnecessary connection between

the two regulations.

The proposed change results in no changes

to plant systems. The proposed change has

no effect on accident conditions or

assumptions. The proposed change does not

affect possible initiating events for accidents

previously evaluated, or any system

functional requirements. Hence, the

proposed change removes an unnecessary tie

between regulations and does not affect plant

operation in any way.

In summary, there is no change in the

probability or consequences of any accident

since the revision of the existing Surveillance

Requirement to reflect the removal of an

unnecessary tie between regulations does not

change the design of the plant, nor the

operational characteristics of any plant

system, nor the procedures by which the

Operators run the plant.

2. The propose change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change removes an

unnecessary tie between regulations. The

objective of the regulation continues to be

met. There are no design changes being made

that would create a new type of accident or

malfunction, and the method and manner of

plant operation remains unchanged.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not involve a

significant reduction in the margin of safety

because they are administrative in nature and

remove an unnecessary tie between

requirements. The proposed change does not

affect any USAR design bases, accident

assumptions. or Technical Specification

Bases. Therefore, the proposed change does

not reduce the margin of safety as defined in

the bases for any TS.

Based upon the above considerations, it

has been concluded that the proposed

changes do not involve significant hazards

considerations.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50-440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: June 9

and 30, 1995

Description of amendment request:

The licensee has requested a one-time

extension of the performance intervals

for certain Technical Specification

Surveillance Requirements (SRs).

Affected SRs include valve testing, and

undervoltage instrumentation testing.

Basis for proposed no significant

hazards considerationdetermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS change requests one-time

only extensions of the surveillance intervals

related to: a) ASME Section XI valve leak

rate, stroke and timing, and position

indication testing; b) Accident Monitoring

Instrumentation related to valve position

indication testing; c) Division 1, 2, and 3

Degraded Voltage and Undervoltage

instrumentation LSFT; and, d) leak rate

testing for hydrostatically tested containment

isolation valves.

Based on the discussion in the License

Amendment Request which shows:

i) The extension of the interval for ASME

Section XI stroke and timing, leak rate

measurement and position indication testing
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42613

requirements are acceptable based on results

of past testing which indicates a margin to TS

limits will be maintained;

ii) The extension of the interval for

Position Indication Calibration as specified

in Table 4.3.7.5-1, Item 17 is acceptable

based on the testing results from the past two

refueling outages that indicate no failures

have occurred:

iii) LSFT interval extension for the

Division 1, 2, and 3 Degraded Voltage and

Undervoltage instrumentation is acceptable

based on the NRC Safety Evaluation Report

(Peach Bottom Atomic Power Plant, Units 2

and 3, dated August 2, 1993) which

supported extension of the interval for LSFT

from 18 to 24 months. This was based on the

small probability of relay or contact failure

relative to mechanical component failure

probability and, therefore, the increase in

LSFT interval represented no significant

change in the overall safety system

unavailability; and,

iv) The extension of the interval for

hydrostatic leak testing of containment

isolation valves is acceptable based on the

consistently low past leak rate data which is

a small percentage of the TS limits.

Therefore, from the above it is shown that

the proposed changes will not significantly

increase the probability of an accident

previously evaluated.

The proposed TS change requests one-time

only extensions of the surveillance intervals

related to TS SR 4.3.3.1, Table 4.3.3.1-1,

Items D.1 and D.2, Division 1, 2, and 3

Degraded Voltage and Undervoltage

instrumentation calibration. [...] extension of

the interval for this instrumentation is

acceptable based on the testing results from

the past two refueling outages. No failures

have occurred which would negate the

assurance that the instrumentation would

function as required for the requested

extended period. Accordingly, the proposed

change will not significantly increase the

probability or consequences of an accident

previously evaluated.

2. The proposed change would not create

the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed TS change requests one-time

extensions of the surveillance intervals for

ASME Section XI valve testing,

instrumentation calibration, instrument

channel LSFT, containment isolation valve

hydrostatic leak rate testing. The proposed

changes do not necessitate a physical

alteration to the plant (no new or different

type of equipment will be installed). In that

the requested extension durations are small

as compared to the overall interval allowed

by TS, NRC and industry evaluations support

extension of LSFT, and past testing results

provide confidence of no effect on equipment

availability by extending the surveillance

interval, the change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS change requests one-time

extensions of the surveillance intervals for

the Division 1, 2, and 3 Undervoltage and

Degraded Voltage instrumentation

calibration. The proposed changes do not

necessitate a physical alteration to the plant

(no new or different type of equipment will

be installed). In that the requested extension

durations are small as compared to the

overall interval allowed by TS and past

testing results provide confidence of no effect

on equipment availability by extending the

surveillance interval, the change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change will not involve a

significant reduction in the margin of safety.

The proposed TS change requests a onetime extension of the surveillance intervals

for ASME Section XI valve testing,

instrumentation calibration, instrument

channel LSFT, and containment isolation

valve hydrostatic leak rate testing. The

proposed changes do not necessitate a

physical alteration to the plant (no new or

different type of equipment will be installed).

In that the requested extension durations are

small as compared to the overall interval

allowed by TS, NRC and industry evaluations

support extension of LSFT, and past testing

results provide confidence of no effect on

equipment availability by extending the

surveillance interval, the change does not

involve a significant reduction in the margin

of safety.

The proposed TS change requests a onetime extension of the surveillance intervals

for the division 1, 2, and 3 Undervoltage and

Degraded Voltage instrumentation

calibration. The proposed changes do not

necessitate a physical alteration to the plant

(no new or different type of equipment will

be installed). In that the requested extension

durations are small as compared to the

overall interval allowed by TS and past

testing results provide confidence of no effect

on equipment availability by extending the

surveillance interval, the change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

July 19, 1995

Description of amendments request:

Amend the Sequoyah Nuclear Plant,

Units 1 and 2 Technical Specification to

incorporate new requirements

associated with steam generator tube

inspections and repair.

Date of publication of individual

notice in the Federal Register: August 1,

1995 (60 FR 39198)

Expiration date of individual notice:

August 31, 1995

Local Public Document Room

Location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

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
42614 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

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 applications for amendments:

December 30, 1993 and July 12, 1994.

The December 30, 1993, application was

supplemented by letters dated

November 30, 1994, May 24, 1995, and

June 21, 1995, and the July 12, 1994,

application was supplemented by letter

dated June 21, 1995.

Brief description of amendments: The

amendments (1) revise the degraded

voltage relay trip setpoint and (2)

enhance the current presentation of the

information regarding the loss-of-voltage

relay setpoint. A time-voltage curve has

been added to the technical

specifications as a more accurate

characterization of the inverse-time

relay response.

Date of issuance: July 21, 1995

Effective date: July 21, 1995, to be

implemented within 45 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 96; Unit 2 -

Amendment No. 84; Unit 3 -

Amendment No. 67

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 8, 1994 and August 17,

1994 (59 FR 29625 and 59 FR 42334)

The November 30, 1994, May 24, 1995,

and June 21, 1995, letters provided

additional clarifying information and

did not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 21, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

February 6, 1995

Brief description of amendment: The

amendment allows the relocation of

cycle-specific core operating limits of

Figure 3.1-1, Shutdown Margin versus

Boron Concentration in Technical

Specification (TS) 3.1.1.2, Shutdown

Margin- Modes 3, 4, and 5, to the plant

Core Operating Limits Report.

Date of issuance: August 1, 1995

Effective date: August 1, 1995

Amendment No. 59

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14017)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

March 30, 1995, as supplemented July 6,

1995. The July 6, 1995, submittal did

not change the initial no significant

hazards consideration determination; it

contained clarifying information only.

Brief description of amendment: The

amendment revises the Emergency

Diesel Generator (EDG) surveillance

requirements contained in TS 3/48.1.1.2

to be consistent with NUREG-1431,

‘‘Standard Technical Specifications,

Westinghouse Plants,’’ and to eliminate

the need for duplicate EDG testing being

performed to satisfy the requirements of

the Station Blackout Rule and the

Maintenance Rule.

Date of issuance: August 1, 1995

Effective date: August 1, 1995

Amendment No.: 60

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20515)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

Docket Nos. STN 50-456 and STN 50-

457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of application for amendments:

June 8, 1995, which superseded the

December 16, 1994, request in its

entirety, and additional correspondence

dated November 30, 1994, April 27, May

5, May 11 and June 23, 1995.

Brief description of amendments: The

amendments revised Figure 3.4-4a

‘‘Nominal PORV Pressure Relief

Setpoint Versus RCS Temperature for

the Cold Overpressure Protection

(LTOP) System’’ in the Braidwood Unit

1’s Technical Specifications. The

revision extends the applicability of

Figure 3.4-4a from 5.37 effective full

power years (EFPY) to 16 EFPY. In

addition, the amendments remove the

638 psig administrative limit line from

the LTOPS curve, because the

appropriate instrument uncertainties

and discharge piping pressure limits

have been incorporated in the new

curve. Finally, the amendments

contains administrative changes to

Figure 3.4-4a and its associated index

page.

Date of issuance: July 24, 1995

Effective date: July 24, 1995

Amendment Nos.: 64 and 64

Facility Operating License Nos. NPF72 and NPF-77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32360).

The June 23, 1995, letter, corrected a

collating error in the June 8, 1995,

submittal and did not change the initial

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 24, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Wilmington Public Library,

201 S. Kankakee Street, Wilmington,

Illinois 60481.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of application for amendments:

March 23, 1994, as supplemented on

July 26, 1994, and subsequently

superseded by a submittal dated
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42615

February 15, 1995. The February 15,

1995, request was supplemented on

February 28, 1995.

Brief description of amendments: The

amendments approve a maximum

moderator temperature coefficient

(MTC) of +7 pcm/°F and relocate

specification of the cycle specific MTC

from the Technical Specifications to the

operating limits report. The staff also

approved the methodology proposed by

the licensee for ensuring that the plants

continue to meet the anticipated

transient without scram (ATWS) rule

(10 CFR 50.62) during operation with

cycle specific MTCs.

Date of issuance: July 27,

1995Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: Byron Units 1 and

2 - 73, 73 and Braidwood Units 1 and

2 - 65, 65

Facility Operating License Nos. NPF37, NPF-66, NPF-72 and NPF-77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18623)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 27, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

July 29, 1992, as supplemented January

14, 1993, February 16, 1993, and May 9,

1995.

Brief description of amendments: The

amendments upgrade the current

custom Technical Specifications for

Dresden and Quad Cities to the

Standard Technical Specifications

contained in NUREG-0123, ‘‘Standard

Technical Specification General Electric

Plants BWR/4.’’ These amendments

upgrade only Section 3/4.3, ‘‘Reactivity

Control.’’

Date of issuance: July 27, 1995

Effective date: Immediately, to be

implemented no later than December

31, 1995, for Dresden Station and June

30, 1996, for Quad Cities Station.

Amendment Nos.: 137, 131, 158, and

154

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 23, 1993 (58 FR 34071)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 27, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

Location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Commonwealth Edison Company,

Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

December 14, 1994

Brief description of amendments: The

amendments revise the surveillance test

intervals and allowed outage times for

certain actuation instrumentation in the

reactor protection, isolation, emergency

core cooling, control rod withdrawal

block, monitoring and feedwater/main

turbine trip systems. The amendments

also include changes to the feedwater/

main turbine trip limiting condition for

operation required actions, several

mode related changes to the nuclear

instrumentation and rod block

specifications, shiftly channel check

requirements for several systems, and

several editorial changes to correct

errors and remove outdated footnotes.

Date of issuance: August 2, 1995

Effective date: Immediately, to be

implemented within 90 days.

Amendment Nos.: 104 and 90

Facility Operating License Nos. NPF11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11128)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 2, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Commonwealth Edison Company,

Docket No. 50-295, Zion Nuclear Power

Station, Unit 1, Lake County, Illinois

Date of application for amendment:

May 17, 1995, as supplemented on June

2, June 16, and July 12, 1995.

Brief description of amendment: The

amendment allows a limited number of

steam generator tubes with roll

transition indications to remain in

service until the September 1995

refueling outage.

Date of issuance: July 26, 1995

Effective date: July 26, 1995

Amendment No.: 167

Facility Operating License No. DPR39: The amendment revises the

Technical Specifications. The June 2,

June 16, and July 12, 1995, submittals

provided additional clarifying

information that did not change the

initial proposed no significant hazards

consideration determination. The

information, however, included changes

to details of the administrative limits

mentioned in the initial proposed no

significant hazards consideration

determination.Public comments

requested as to proposed no significant

hazards consideration determination:

Yes (60 FR 27798). This notice provided

an opportunity to submit comments on

the Commission’s proposed no

significant hazards consideration

determination. No comments have been

received. The notice also provided for

an opportunity to request a hearing by

June 26, 1995, but indicated that if the

Commission makes a final no significant

hazards consideration determination

any such hearing would take place after

issuance of the amendment. The

Commission’s related evaluation of the

amendments, finding of exigent

circumstances and final no significant

hazards consideration determination is

contained in a Safety Evaluation dated

July 26, 1995.

Local Public Document Room

Location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Consumers Power Company, Docket

No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of application for amendment:

December 15, 1994

Brief description of amendment: The

amendment revises Technical

Specification 11.3.1.5 ACTION a. to

eliminate the need to demonstrate that

the actuation circuitry of the unaffected

reactor depressurization system

channels is operable. In addition, the

amendment makes an editorial change

to correct a typographical error.

Date of issuance: July 28, 1995

Effective date: July 28, 1995

Amendment No.: 115

Facility Operating License No. DPR-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20516)

The Commission’s related evaluation of

the amendment is contained in a Safety
42616 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

Evaluation dated July 28, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770.

Consumers Power Company, Docket

No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of application for amendment:

March 4, 1993, as revised April 14,

1993, as supplemented April 19 and

May 31, 1995

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to conform to the

wording of the revised 10 CFR Part 20,

‘‘Standards for Protection Against

Radiation,’’ and to reflect a separation of

chemistry and radiation protection

responsibilities.

Date of issuance: August 2, 1995

Effective date: August 2, 1995

Amendment No.: 16

Facility Operating License No. DPR-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 12, 1993 (58 FR 28053),

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

The supplemental submittals were

noticed on June 21, 1995 (60 FR 32361).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation datedNo significant hazards

consideration comments received: No.

Local Public Document Room

Location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

April 7, 1994, as supplementedApril 27,

1995.

Brief description of amendment: This

amendment relocates certain Technical

Specifications (TS) that contain fuel

cycle-specific parameter limits that

change with core reloads to a Core

Operating Limits Report. TS bases have

also been revised to refer to limits

relocated to the COLR. A portion of the

amendment request was denied. A

separate Notice of Denial of Amendment

has been sent to the Federal Register for

publication.

Date of issuance: July 26, 1995

Effective date: July 26, 1995

Amendment No.: 169

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27053)

The April 27, 1995, submittal provided

clarifying information which was within

the scope of the initial application and

did not affect the staff’s initial proposed

no significant hazards consideration

findings. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 26, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

Location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

April 12, 1995

Brief description of amendments: The

amendments delete Technical

Specification (TS) 3/4.3.4, ‘‘Turbine

Overspeed Protection,’’ and its

associated Bases. The deletion of TS 3/

4.3.4 and its Bases provides Duke Power

Company the flexibility to implement

the manufacturer’s recommendations for

turbine steam valve surveillance test

requirements. These test requirements

will be contained in the Selected

Licensee Commitment Manual.

Date of issuance: July 21, 1995

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance

Amendment Nos.: 131 and 125

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

the Technical Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32361)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 21, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

Location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

January 18, 1995.

Brief description of amendments: The

amendments relocate the requirements

for the seismic instrumentation,

meteorological instrumentation, and

loose-part detection system, and the

associated Bases and surveillance

requirements, from the TS to the

Selected Licensee Commitment Manual

(Chapter 16 of the FSAR). This will

allow future changes to these controls to

be performed under the provisions of 10

CFR 50.59. No changes are being made

to the technical content of the affected

TS pages.

Date of issuance: July 24, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 132 and 126

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

the Technical Specifications.

Date of initial notice in Federal

Register: May 10, 1995 (60 FR 24910)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 24, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

April 12, 1995

Brief description of amendments: The

amendments delete Technical

Specification (TS) 3/4.3.4, ‘‘Turbine

Overspeed Protection,’’ and its

associated Bases. The deletion of TS 3/

4.3.4 and its associated Bases provides

Duke Power Company the flexibility to

implement the manufacturer’s

recommendations for turbine steam

valve surveillance test requirements.

These test requirements will be

contained in the Selected Licensee

Commitment (SLC) Manual. The SLC

Manual is Chapter 16 of the Updated

Final Safety Analysis Report.

Date of issuance: August 2, 1995

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance

Amendment Nos.: 156 and 138

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

Technical Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32362)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 2, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

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:

September 28, 1994, as supplemented
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42617

by letters dated May 3 and June 14,

1995.

Brief description of amendments: The

amendments revise Technical

Specification Tables 3.3-3, 3.3-4, 3.3-5,

and 4.3-2 of the Engineered Safety

Features Actuation System

Instrumentation tables to update the

‘‘Loss of Power’’ function.

Date of issuance: August 2, 1995

Effective date: As of the date of

issuance to be implemented within 60

days, or 60 days after the completion

date of the Unit 2 modification,

whichever is later.

Amendment Nos.: 157 and 139

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

Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65811) The May 3 and June 14, 1995,

letters provided clarifying information

that did not change the scope of the

September 28, 1994, application and the

initial proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated August 2, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

January 18, 1995

Brief description of amendments: The

amendments delete selected Technical

Specification (TS) requirements related

to instrumentation from the TS, and

relocate them to the Selected Licensee

Commitment (SLC) Manual, with their

associated Bases and surveillance

requirements. No changes are being

made to the technical content of the

affected TS pages. Future changes to the

SLC Manual (Chapter 16 of the Final

Safety Analysis Report) will be

controlled by the provisions of 10 CFR

50.59. The relocated requirements

include the following:

TS 3/4.3.3.3, Seismic Instrumentation

TS 3/4.3.3.4, Meteorological

Instrumentation

TS 3/4.10, Loose-Part Detection

System

Date of issuance: August 2, 1995

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance

Amendment Nos.: 158 and 140

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

Technical Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11132)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 2, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

February 4, 1994, as supplemented June

29, 1995.

Brief description of amendments:

These amendments modify the

Technical Specifications (TSs) related to

containment air locks (TSs 1.8, 3/4.6.1.1

and 3/4.6.1.3) and associated Bases to

make them as close to the NRC’s

Improved Standard Technical

Specifications (NUREG-1431) as the

plant-specific design will permit. The

changes in TS 3/4.6.1.1 and 3/4.6.1.3

modify surveillance requirements and

limiting conditions for operation and

effect numerous administrative and

format changes.

Date of issuance: July 26, 1995

Effective date: Units 1 and 2, as of the

date of issuance and shall be

implemented within 60 days.

Amendment Nos.: 190 and 72

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Units 1 and 2 Technical

Specifications, and the Unit 2 License.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37070)

The June 29, 1995 letter did not change

the original no significant hazards

consideration determination or expand

the scope of the July 20, 1994 Federal

Register notice.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

July 26, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

Location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request: May 12,

1995

Brief description of amendment: The

amendment removed the specific

scheduling requirements for Type A

containment leakage rate tests from the

Technical Specifications for Waterford 3

and replaced these requirements with a

requirement to perform Type A, testing

in accordance with Appendix J to 10

CFR Part 50. The proposed changes

adopt the wording for primary

containment integrated leak rate testing

that is consistent with the requirements

of the Combustion Engineering

Improved Standard Technical

Specifications (NUREG 1432). The

proposed changes also include several

administrative changes.

Date of issuance: August 3, 1995

Effective date: August 3, 1995, to be

implemented within 60 days of

issuance.

Amendment No.: 110

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29876)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 3, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

October 13, 1994, as supplemented by

letters dated January 13 and May 4,

1995.

Brief description of amendments: The

amendments revise the Technical

Specifications to lower the anticipated

transient without scram-recirculation

pump trip (ATWS-RPT) setpoint by

approximately 2 feet 2 inches to

minimize the potential for RPTs

following reactor scram, and allow

restarting the recirculation pump

following an RPT when the temperature

differential between the coolant at the

reactor bottom head and the reactor

steam dome cannot be obtained,

provided certain conditions are met.

Date of issuance: July 21, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment Nos.: 196 and 136

Facility Operating License Nos. DPR57 and NPF-5. Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR
42618 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

65813). The January 13 and May 4,

1995, letters provided clarifying

information that did not change the

scope of the October 13, 1994,

application and initial proposed no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

July 21, 1995. No significant hazards

consideration comments received: No

Local Public Document Room

Location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513.

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of application for amendment:

June 1, 1995

Brief description of amendment: The

amendment revises the TMI-1 Technical

Specifications to allow the use of two

zirconium-based advanced fuel rod

cladding materials manufactured by the

Babcock & Wilcox Fuel Company.

Date of issuance: July 24, 1995

Date of issuance: July 24, 1995

Effective date: July 24, 1995

Amendment No.: 194

Facility Operating License No. DPR50. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32366)

The Commission’s related evaluation of

this amendment is contained in a Safety

Evaluation dated July 24, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: May 13,

1993 as supplemented by letter dated

January 31, 1995

Brief description of amendment: The

amendment revises Attachment 3 of the

license conditions to remove several

license conditions pertaining to the

Division I and II Transamerica Delaval,

Inc. emergency diesel generators. The

conditions pertain to engine overhaul

frequency, maintenance and

surveillance program, and inspection of

crankshafts, cylinder heads, engine

block, and turbochargers.

Date of issuance: July 25, 1995

Effective date: July 25, 1995

Amendment No.: 82

Facility Operating License No. NPF47. The amendment revised the

operating license.

Date of initial notice in Federal

Register: August 4, 1993 (58 FR 41505)

The additional information contained in

the supplemental letter dated January

31, 1995, was clarifying in nature and

thus, within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 25, 1995.No

significant hazards consideration

comments received. No.

Local Public Document Room

Location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: April 27,

1995, as supplemented by letters dated

May 4 and 25, 1995.

Brief description of amendments: The

amendments revised the tables

associated with Technical

Specifications (TSs) 3/4.3.3.5, Remote

Shutdown System, to eliminate the

requirement for core exit thermocouples

(CETs). The amendments also revised

the tables associated with TS 3/4.3.3.6,

Accident Monitoring Instrumentation,

to require two operable channels of

CETs, where each channel is required to

have at least two operable CETs per core

quadrant. Each channel is also required

to have at least four operable CETs in at

least one quadrant to support the

operability of the subcooling margin

monitors.

Date of issuance: July 24, 1995

Effective date: July 24, 1995

Amendment Nos.: Unit 1 -

Amendment No. 77; Unit 2 -

Amendment No. 66

Facility Operating License Nos. NPF76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32366)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 24, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 2,

1995

Brief description of amendments: The

amendments revised Technical

Specifications 3.4.2.2. and 3.7.1.1 (Table

3.7-2) by relaxing the lift setting

tolerances of the pressurizer safety

valves from plus or minus 1% to plus

or minus 2% and the main steam safety

valves from plus or minus 1% to plus

or minus 3%, respectively. In addition,

a footnote was added to require that the

pressurizer safety valves and main

steam safety valves setpoint tolerances

be restored to within plus or minus 1%

whenever a lift setting is determined to

be outside plus or minus 1% following

valve testing.

Date of issuance: July 25, 1995

Effective date: July 25, 1995, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 78; Unit 2 -

Amendment No. 67

Facility Operating License Nos. NPF76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29877)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated July 25, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

March 7, 1995, as supplemented on June

7, 1995.

Brief description of amendment: The

amendment adds an Exception to

Technical Specifications 3.6.A and

3.6.C. The Exception permits reduced

component cooling water flow for short

periods of time, while component

cooling water heat exchangers are

shifted.

Date of issuance: July 24, 1995
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42619

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 151

Facility Operating License No. DPR36: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 10, 1995 (60 FR 24911)

The June 7, 1995, submittal 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

July 24, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

Location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME

04578.

Northeast Nuclear Energy Company,

Docket No. 50-245, Millstone Nuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

May 24, 1995

Brief description of amendment: The

amendment permits an individual who

does not have a current senior reactor

operator (SRO) license for Millstone

Unit 1 to hold the Operations Manager

position. In this case, the Operations

Manager position would require the

individual to have previously held an

SRO license at a boiling water reactor

and the individual serving in the

capacity of the Assistant Operations

Manager to hold a current SRO license

for Millstone Unit 1. In addition, the

amendment renumbers the applicable

sections.

Date of issuance: July 24, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 83

Facility Operating License No. DPR21. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32370)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 24, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Northeast Nuclear Energy Company,

Docket No. 50-245, Millstone Nuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

April 18, 1995

Brief description of amendment: The

amendment allows the use of the ANSI/

ANS 5.1-1979 decay heat model for the

post-loss of coolant accident

containment cooling analysis.

Date of issuance: July 24, 1995

Effective date: As of the date of

issuance to be implemented

immediately.

Amendment No.: 84

Facility Operating License No. DPR21. Amendment revised the license.

Date of initial notice in Federal

Register: May 10, 1995 (60 FR 24911).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 24, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

April 28, 1995

Brief description of amendment: The

amendment revises the diesel generator

fuel oil testing that is performed on new

fuel prior to the addition of new fuel to

the storage tank.

Date of issuance: July 26, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 118

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29881)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 26, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

Location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

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

Atomic Power Station,Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments:

November 14, 1994 as supplemented by

letter dated April 10, 1995.

Brief description of amendments:

These amendments relocate Nuclear

Review Board (NRB) review

requirements, Independent Safety

Engineering Group (ISEG) requirements,

and certain review and audit

requirements from the TS to the Peach

Bottom Quality Assurance Program.

Date of issuance: July 25, 1995

Effective date: July 25, 1995

Amendments Nos.: 208 and 212

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65822) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 25, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

Location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

July 27, 1994, as supplemented May 26,

July 10, and July 25, 1995

Brief description of amendment: This

amendment revises the Allowed Out-ofService Times (AOTs) for Inoperable

Station Service Water System (SSWS)

pumps, inoperable safety Auxiliaries

Cooling System (SACS) pumps, and

inoperable Emergency Diesel Generators

(EDGs). In addition, this amendment

also allows on-line maintenance of the

EDGs.

Date of issuance: August 1, 1995

Effective date: August 1, 1995

Amendment No.: 75

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR 45033)

The supplemental letters did not change

the original no significant hazards

consideration determination nor the

original Federal Register notice. The
42620 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

April 25, 1994, as supplemented July

24, 1995

Brief description of amendment: This

amendment eliminates the requirement

from the Hope Creek Technical

Specifications to perform Type C leak

rate tests, in accordance with 10 CFR

Part 50, Appendix J, of identified

containment isolation valves that

penetrate the primary containment and

terminate below the minimum water

level in the suppression chamber

(torus). The valves are still subject to

testing in accordance with the American

Society of Mechanical Engineers Boiler

and Pressure Vessel Code.

Date of issuance: August 1, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 76

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 8, 1994 (59 FR 29632)

The supplemental letter did not change

the original no significant hazards

consideration determination nor the

original Federal Register notice.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

April 18, 1995

Brief description of amendments: The

amendments delete the quarterly leak

rate test for the containment pressurevacuum relief valves that is currently

required because of the valves’ resilient

seat material. The changes are being

made to accommodate replacement of

the resilient valve seat material with a

hard seat (metal-to-metal) design. The

valves would remain in the 10 CFR Part

50, Appendix J, Type C leak rate test

program.

Date of issuance: August 1, 1995

Effective date: Unit 1, As of the date

of issuance, to be implemented prior to

restart following the twelfth refueling

outage; Unit 2, As of the date of

issuance, to be implemented prior to

restart following the current refueling

outage.

Amendment Nos.: 172 and 153

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 23, 1995 (60 FR 27342)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Salem Free Public Library,

112 West Broadway, Salem, New Jersey

08079.

Southern 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:

August 26, 1994

Brief description of amendments:

These amendments revise Technical

Specification 3/4.7.5, ‘‘Control Room

Emergency Air Cleanup System,’’ to

provide an exception to Limiting

Condition for Operation 3.0.4 for Modes

5 and 6 and for a defueled

configuration. These amendments also

add the applicability statement ‘‘or

during movement of irradiated fuel

assemblies.’’

Date of issuance: July 26, 1995

Effective date: July 26, 1995

Amendment Nos.: Unit 2 -

Amendment No. 123; Unit 3 -

Amendment No. 112

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55891) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 26, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

Location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

December 16, 1994; supplemented July

19, 1995 (TS 94-06)

Brief description of amendments: The

amendments replace the present

Auxiliary Feedwater system

Specification 3/4.7.1.2 with new

specifications that are modeled after the

Westinghouse Standard Technical

Specifications.

Date of issuance: August 2, 1995

Effective date: August 2, 1995

Amendment Nos.: 206 and 196

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

technical specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6309)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 2, 1995.No

significant hazards consideration

comments received: None

Local Public Document Room

Location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Virginia Electric and Power Company,

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

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia.

Date of application for amendments:

November 29, 1994

Brief description of amendments:

These amendments allow the use of

ZIRLO, a new zirconium-based alloy, as

a fuel cladding material.

Date of issuance: July 27, 1995

Effective date: July 27, 1995

Amendment Nos.: 202 and 202

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 508)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 27, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

Location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Notice Of Issuance Of Amendments to

facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the
Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices 42621

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

September 15, 1995, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such
42622 Federal Register / Vol. 60, No. 158 / Wednesday, August 16, 1995 / Notices

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

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

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of application for amendment:

July 28, 1995

Brief description of amendment: This

amendment deletes the portion of

License Condition 2.C.(1) that references

Attachment 1. Attachment 1 requires

the pump in the keepwarm system on

the emergency diesel generator to satisfy

the requirements of the American

Society of Mechanical Engineers Code,

Section III, Class 3.

Date of issuance: August 3,

1995I11Effective date: August 3, 1995

Amendment No.: 88

Facility Operating License No. NPF42: The amendment revised the

operating license.Public comments

requested as to proposed no significant

hazards consideration: No.The

Commission’s related evaluation of the

amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated August 3, 1995.

Local Public Document Room

Location: 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

Dated at Rockville, Maryland, this 16th day

of August 1995.

For The Nuclear Regulatory Commission

Jack W. Roe,

Director, Division of Reactor Projects - III/

IV Office of Nuclear Reactor Regulation

[Doc. 95–20122 Filed 8–15–95; 8:45 am]

BILLING CODE 7590–01–F

Use of Probabilistic Risk Assessment

Methods in Nuclear Regulatory

Activities; Final Policy Statement

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final policy statement.

SUMMARY: This statement presents the

policy that the Nuclear Regulatory

Commission (NRC) will follow in the

use of probabilistic risk assessment

(PRA) methods in nuclear regulatory

matters. The Commission believes that

an overall policy on the use of PRA

methods in nuclear regulatory activities

should be established so that the many

potential applications of PRA can be

implemented in a consistent and

predictable manner that would promote

regulatory stability and efficiency. In

addition, the Commission believes that

the use of PRA technology in NRC

regulatory activities should be increased

to the extent supported by the state-ofthe-art in PRA methods and data and in

a manner that complements the NRC’s

deterministic approach. The pertinent

comments received from the published

draft policy statement are reflected in

this final policy statement. This policy

statement will be implemented through

the execution of the NRC’s PRA

Implementation Plan.

EFFECTIVE DATE: August 16, 1995.

ADDRESSES: The proposed policy

statement and the comments received

may be examined at: NRC Public

Document Room, 2120 L Street, NW.

(Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT:

Anthony Hsia, Office of Nuclear Reactor

Regulation, U.S. Nuclear Regulatory

Commission, Washington, DC 20555.

Telephone (301) 415–1075.

SUPPLEMENTARY INFORMATION:

I. Background.

II. Summary of Public Comments and NRC

Responses.

III. Deterministic and Probabilistic

Approaches to Regulation.

IV. The Commission Policy.

V. Availability of Documents.

I. Background

The NRC has generally regulated the

use of nuclear material based on

deterministic approaches. Deterministic

approaches to regulation consider a set

of challenges to safety and determine

how those challenges should be

mitigated. A probabilistic approach to

regulation enhances and extends this

traditional, deterministic approach, by:

(1) Allowing consideration of a broader

set of potential challenges to safety, (2)

providing a logical means for

prioritizing these challenges based on

risk significance, and (3) allowing

consideration of a broader set of

resources to defend against these

challenges.

Until the accident at Three Mile

Island (TMI) in 1979, the Atomic Energy

Commission (now the NRC), only used

probabilistic criteria in certain

specialized areas of licensing reviews.

For example, human-made hazards (e.g.,

nearby hazardous materials and aircraft)

and natural hazards (e.g., tornadoes,

floods, and earthquakes) were typically

addressed in terms of probabilistic

arguments and initiating frequencies to

assess site suitability. The Standard

Review Plan (NUREG–0800) for

licensing reactors and some of the

Regulatory Guides supporting NUREG–

0800 provided review and evaluation

guidance with respect to these

probabilistic considerations.

The TMI accident substantially

changed the character of the analysis of

severe accidents worldwide. It led to a

substantial research program on severe

accident phenomenology. In addition,