Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices 3497

Testing of Safety-Related Logic

Circuits; Issued

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of Issuance.

SUMMARY: The Nuclear Regulatory

Commission (NRC) has issued Generic

Letter 96–01 to notify licensees of

nuclear power reactors about problems

with testing of safety-related logic

circuits, request that licensees

implement certain actions, and require

that all licensees submit a written

response. This generic letter is available

in the Public Document Rooms under

accession number 9601050193.

DATES: The generic letter was issued on

January 10, 1996.

ADDRESSEES: Not applicable.

FOR FURTHER INFORMATION CONTACT:

Hukam C. Garg at (301) 415–2929.

SUPPLEMENTARY INFORMATION: None.

Dated at Rockville, Maryland, this 23rd day

of January, 1996.

For the Nuclear Regulatory Commission.

Dennis M. Crutchfield,

Director, Division of Reactor Program

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 96–1871 Filed 1–30–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from January 5,

1996, through January 19, 1996. The last

biweekly notice was published on

January 22, 1996.

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

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 March 1, 1996, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.
3498 Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, 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.

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: January

12, 1996

Description of amendment request:

Compliance with 10 CFR Part 50,

Appendix J, provides assurance that the

primary containment, including those

systems and components that penetrate

the primary containment, do not exceed

the allowable leakage rate values

specified in the Technical

Specifications and Bases. The allowable

leakage rate is determined so that the

leakage assumed in the safety analyses

is not exceeded.

On February 4, 1992, the NRC

published a notice in the Federal

Register (57 FR 4166) discussing a

planned initiative to begin eliminating

requirements marginal to safety that

impose a significant regulatory burden.

Appendix J to 10 CFR Part 50, ‘‘Primary

Containment Leakage Testing for WaterCooled Power Reactors,’’ was

considered for this initiative and the

staff undertook a study of possible

changes to this regulation. The study

examined the previous performance

history of domestic containments and

examined the effect on risk of a revision

to the requirements of Appendix J. The

results of this study are reported in

NUREG–1493, ‘‘Performance-Based

Leak-Test Program.’’

Based on the results of this study, the

staff developed a performance based

approach to containment leakage rate

testing. On September 12, 1995, the

NRC approved issuance of this revision

to 10 CFR Part 50, Appendix J, which

was subsequently published in the

Federal Register on September 26, 1995,

and became effective on October 26,

1995. The revision added Option B

‘‘Performance-Based Requirements’’ to

Appendix J to allow licensees to

voluntarily replace the prescriptive

testing requirements of Appendix J with

testing requirements based on both

overall and individual component

leakage rate performance.

Regulatory Guide 1.163,

‘‘Performance-Based Containment Leak

Test Program,’’ was developed as a

method acceptable to the staff for

implementing Option B. Accordingly,

the licensee has submitted, in its

application dated January 12, 1996,

proposed changes to the TS to

implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leakage-Test Program.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1The proposed change will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Containment leak rate testing is not an

initiator of any accident; the proposed

change does not affect reactor operations or

accident analysis, and has no significant

radiological consequences. Therefore, this

proposed change will not involve an increase

in the probability or consequences of any

previously-evaluated accident.

2. The proposed change will not create the

possibility of any new accident not

previously evaluated.

The proposed change does not affect

normal plant operations or configuration, nor

does it affect leak rate test methods. The test

history at Catawba (no ILRT [integrated leak

rate test] failures) provides continued

assurance of the leak tightness of the

containment structure.

3. There is no significant reduction in a

margin of safety.
Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices 3499

The proposed changes are based on NRCaccepted provisions, and maintain necessary

levels of reliability of containment integrity.

The performanced-based approach to leakage

rate testing recognizes that historically good

results of containment testing provide

appropriate assurance of future containment

integrity; this supports the conclusion that

the impact on the health and safety of the

public as a result of extended test intervals

is negligible.

Based on the above, no significant hazards

consideration is created by the proposed

change.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

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:

December 27, 1995

Description of amendment request:

The proposed amendments would

modify Tables 3.3–11 and 4.3–7 of

Beaver Valley Power Station Unit Nos.

1 and 2 (BVPS–1 and BVPS–2)

Technical Specification (TS) 3.3.3.8

such that only one valve position

indication system for the power

operated relief valves and safety valves

is required to be operable. The licensee

stated that the proposed amendments

would then be consistent with the

NRC’s Improved Standard Technical

Specifications, NUREG–1431, Revision

1, and with the guidance of Regulatory

Guide 1.97, NUREG–0578, and NUREG–

0737.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change involves

instrumentation which is redundant in

monitoring the position of valves and, as

such, does not influence the potential for an

initiating event involving the power operated

relief valves (PORVs) or the safety valves

(SVs). Implementation of these changes will

reduce the potential for challenges to the

plant due to a potential shutdown which

should not be necessary due to the restrictive

nature of having unnecessary redundant

position indication in the technical

specification. By deleting the Unit No. 1

technical specification operability

requirements for the PORV acoustic

detectors, and by deleting, on both units, the

technical specification operability

requirements for the SV temperature detector

position indicators, the potential for

unnecessary shutdowns is reduced. When

inoperable, the PORV acoustic detectors and

the SV temperature detectors presently

invoke an unnecessary action statement as

another fully qualified safety-related position

indication system exists to provide

indication. The proposed change modifies

Specification 3.3.3.8 actions and surveillance

requirements, but does not affect the BASES.

The remaining instrumentation on these

tables [3.3–11 and 4.3–7] will be unaffected.

The remaining position indication systems

for the PORVs and SVs are fully qualified

and satisfy regulatory criteria for post

accident monitoring of valve position. These

changes do not affect the ability to satisfy

analysis assumptions regarding operation of

the PORVs and SVs. They do not affect the

ability to continue to meet the guidance of

Regulatory Guide 1.97, the post Three Mile

Island criteria contained in NUREG 0578 and

NUREG 0737, and reflect the guidance

provided in NUREG 1431, ‘‘Improved

Standard Technical Specifications’’ (ISTS).

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 Updated

Final Safety Analysis Report (UFSAR).

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change will reduce the

potential to challenge safety systems due to

eliminating the potential for unnecessary

plant shutdowns. The proposed changes are

limited to PORV and SV position indication

and do not involve any physical changes to

the PORVs or SVs or their setpoints. These

changes do not delete any design basis

accident functions previously provided by

the PORVs or SVs nor has the probability of

inadvertent opening been increased.

Accordingly, no new 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 eliminate a degree of

equipment redundancy and is consistent

with the Improved Standard Technical

Specifications (ISTS). The Unit No. 1

specification presently requires operability of

both redundant PORV position indication

systems and the primary and backup SV

position indication systems. The Unit No. 2

specification also requires operability of the

primary and backup SV position indication

systems. These changes will potentially

eliminate some challenges and potential

unnecessary shutdowns by eliminating

equipment determined to be no longer

necessary. Only one safety-related position

indication system is necessary to satisfy

regulatory criteria; 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

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request:

December 22, 1995

Description of amendment request:

The proposed amendment would revise

the Duane Arnold Energy Center (DAEC)

Technical Specifications (TS) Sections

3.7.A and 4.7.A, ‘‘Primary

Containment,’’ by deleting information

also contained in 10 CFR Part 50,

Appendix J, Option A and incorporating

references to the Primary Containment

Leakage Rate Testing Program. These

changes will allow the use of the

performance based option of

containment leak testing. The request

also adds Operability and Surveillance

Requirements (SRs) for the drywell air

lock. Minor administrative changes are

also made. These changes are consistent

with comparable specifications in the

Improved Standard Technical

Specifications (ITS), NUREG–1433. In

addition to the licensee’s proposed

revision to the DAEC TS, the staff will

be executing administrative changes and

corrections to the TS Bases, as

submitted in letters(2) dated February

13, 1995. Sections that will be changed

or corrected are Section 1.2, Bases;

Section 2.2, Bases Reactor Coolant

System Integrity; Section 3.2, Bases;

Section 3.7.H/4.7.H, Bases Containment

Atmosphere Dilution; and Section 3.7.I/

4.7.I, Bases Oxygen Concentration.

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
3500 Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices

consideration, which is presented

below:

. The proposed revision does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Information contained in 10 CFR 50,

Appendix J was deleted and references to the

Primary Containment Leakage Rate Testing

Program were added. These are

administrative changes to allow the use of

performance-based containment leakage

testing methods. The containment testing

program will conform with the requirements

of Option B of 10 CFR Part 50, Appendix J

and approved exemptions. The performance

of the leakage tests themselves is not an input

or consideration in any accident previously

evaluated, thus the proposed change will not

increase the probability of any such accident

occurring. The same operability requirements

remain for the primary containment,

therefore the consequences of an accident are

not significantly increased.

Drywell air lock operability and

surveillance requirements were added.

Actions for one air lock door inoperable have

been added consistent with the ITS. In

addition, notes have been added to allow

entry and exit to perform repairs of the air

lock components and to explain that the

previous overall leak test is not invalidated

by an inoperable door. This change

represents an additional restriction on plant

operation, since the previous condition of

one air lock door inoperable did not require

any actions to be taken. A requirement to

verify proper operation of interlock

mechanism was also added. This will ensure

that one door is always closed which

maintains primary containment integrity.

The addition of these new drywell air lock

requirements provides more stringent

provisions than previously existed in the

[current Technical Specifications]. The more

stringent requirements will not result in

operation that will increase the probability of

initiating an analyzed event. If anything, the

new requirements may decrease the

probability or consequences of an analyzed

event by incorporating the more restrictive

changes discussed above. These changes will

not alter assumptions relative to mitigation of

an accident or transient event. The more

restrictive requirements will not alter the

operation of process variables, structures,

systems, or components as described in the

safety analyses.

The TS revision includes the relocation of

certain requirements from the current

technical specification (CTS) to licensee

controlled documents. CTS 4.7.A.1.e

contains a requirement to replace the T-ring

inflatable seals for the 18 inch purge valves

every four years. This provision is not in the

ITS as it is a maintenance issue and not a

surveillance for operability. CTS 4.7.A.1.e

also contains a requirement to verify (during

Type C testing) that the mechanical

modification which limits the maximum

opening angle for the 18 inch purge valves

is intact. The ITS only requires this

surveillance if the mechanical modification

is not permanent. At DAEC, the 18 inch

purge valves are permanently blocked to

restrict opening to 30°. These CTS provisions

will be relocated to plant procedures. Any

changes to these relocated requirements will

require an evaluation in accordance with 10

CFR 50.59. CTS 4.7.A.1.a and 4.7.A.1.d

contain some procedural details that are not

contained in Appendix J. These details will

also be relocated to plant procedures,

consistent with the ITS. Since any changes to

these licensee controlled documents will be

evaluated in accordance with 10 CFR 50.59,

no significant increase in the probability or

consequences of an accident previously

evaluated will be allowed.

The proposed revision does not involve

any change to the configuration or method of

operation of any plant equipment that is used

to mitigate the consequences of an accident,

nor does it affect any assumptions or

conditions in the accident analysis. The

proposed revision does not degrade any

existing plant programs, nor modify any

functions of safety related systems or

accident mitigation functions previously

credited at the DAEC. The proposed changes

do not impact initiators of analyzed events.

They also do not impact the assumed

mitigation of accidents or transient events.

These TS changes will not alter assumptions

made in the safety analysis and licensing

basis.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed revision does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Deleting information from the TS which is

contained in 10 CFR 50, Appendix J and

adding references to the Primary

Containment Leakage Rate Testing Program

are purely administrative changes to allow

the use of performance-based containment

leakage testing methods. The containment

testing program will conform with the

requirements of Option B of 10 CFR Part 50,

Appendix J and approved exemptions. The

use of Option B will maintain the

containment safety functions as a barrier to

the release of radioactivity to the

environment.

The proposed revision does not make any

physical or operational changes to existing

plant systems or components, nor does it

alter any plant parameters, revise any safety

limit setpoint, or provide any new release

pathways. The proposed revision does not

change any transient responses assumed in

the Design Bases of the plant.

The proposed changes which relocate

requirements to licensee controlled

documents will not alter the plant

configuration (no new or different type of

equipment will be installed) or change the

methods governing normal plant operation.

These changes will not alter assumptions

made in the safety analysis or licensing basis.

The proposed changes which add more

restrictive requirements to the CTS will not

alter the plant configuration (no new or

different type of equipment will be installed)

or change the methods governing normal

plant operation. These changes do impose

different requirements. However, they are

consistent with assumptions made in the

safety analyses.

Therefore, the revision does not create the

possibility of a new or different kind of

accident previously evaluated.

3. The proposed revision will not

significantly reduce any margin of

safety.

Deleting information from the TS which is

contained in 10 CFR 50, Appendix J and

adding references to the Primary

Containment Leakage Rate Testing Program

do not involve a significant reduction in the

margin of safety. These changes 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 containment

testing program will conform to the

requirements of Option B of 10 CFR Part 50,

Appendix J and approved exemptions. The

use of Option B will maintain the

containment safety functions as a barrier to

the release of radioactivity to the

environment.

The proposed revision does not require any

modifications to existing plant systems or

equipment, safety limit settings, or

parameters utilized in the licensing bases for

the safety analysis. The proposed revision

does not change any safety analysis or any

accident mitigation action for which DAEC

has previously taken credit. The proposed

changes do not involve any technical

changes; they have no impact on any safety

analysis assumptions. The addition of new

requirements either increases or does not

affect the margin of safety.

The proposed changes that relocate

requirements from the CTS to licensee

controlled documents will not reduce a

margin of safety since they have no impact

on any safety analysis assumptions. In

addition, the requirements to be relocated

from the CTS to the licensee controlled

document are unchanged. Since any future

changes to this licensee controlled document

will be evaluated in accordance with the

requirements of 10 CFR 50.59, no significant

reduction in a margin of safety will be

allowed.

The proposed changes are consistent with

NUREG–1433, which was approved by the

NRC Staff. The changes are also consistent

with NRC guidance provided for the

implementation of Option B. The change

controls for proposed relocated details and

requirements are acceptable. Therefore,

revising the TS to reflect the NRC accepted

level of detail and requirements ensures that

there is no reduction in a margin of safety.

Therefore, the proposed revision will not

significantly reduce any margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401
Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices 3501

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis, &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869

NRC Project Director: Gail H. Marcus

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No.

50–461, Clinton Power Station, Unit No.

1, DeWitt County, Illinois

Date of amendment request:

December 14, 1995

Description of amendment request:

The proposed amendment would

modify Technical Specifications 3.3.1.1,

‘‘Reactor Protection System (RPS)

Instrumentation,’’ and 3.3.6.1, ‘‘Primary

Containment and Drywell Isolation

Instrumentation,’’ to eliminate periodic

response time testing of selected analog

trip modules (ATMs). This request is

supported by analyses prepared by the

Boiling Water Reactor Owners’ Group

topical report NEDO–32291, ‘‘System

Analyses for Elimination of Selected

Response Time Testing Requirements,’’

which demonstrate that other periodic

tests required by technical

specifications, such as channel

calibrations, channel functional tests

and logic system functional tests, are

adequate to ensure ATM response times

remain within acceptable limits.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

(1) The purpose of the proposed Technical

Specification (TS) change is to eliminate

response time testing requirements for

selected analog trip modules (ATMs) in the

Reactor Protection System (RPS) and the

main steam isolation valve (MSIV) isolation

actuation instrumentation. The Boiling Water

Reactor Owners’ Group (BWROG) has

completed an evaluation which demonstrates

that response time testing is redundant to the

other TS-required testing. These other tests,

in conjunction with actions taken in response

to NRC Bulletin 90–01, ‘‘Loss of Fill-Oil in

Transmitters Manufactured by Rosemount,’’

and Supplement 1, are sufficient to identify

failure modes or degradations in instrument

response time and ensure operation of the

associated systems within acceptable limits.

There are no known failure modes that can

be detected by response time testing that

cannot also be detected by other TS-required

testing. This evaluation was documented in

NEDO–32291, ‘‘System Analyses for

Elimination of Selected Response Time

Testing Requirements,’’ January 1994. Illinois

Power (IP) has confirmed the applicability of

this evaluation to Clinton Power Station

(CPS). In addition, IP has completed the

actions identified in the NRC staff’s safety

evaluation of NEDO–32291.

Because of the continued application of

other existing TS-required tests 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 changes do not affect the capability

of the associated systems to perform their

intended function within their required

response time, nor do the proposed changes

themselves affect the operation of any

equipment. As a result, IP has concluded that

the proposed changes do not involve a

significant increase in the probability or the

consequences of an accident previously

evaluated.

(2) The proposed changes only apply to the

testing requirements for ATMs in the systems

identified above and do not result in any

physical change to these or other components

or their operation. As a result, no new failure

modes are introduced. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) The current TS-required response times

are based on the maximum values assumed

in the plant safety analyses. These analyses

conservatively establish the margin of safety.

As described above, the proposed changes do

not affect the capability of the associated

systems to perform their intended function

within the allowed response time used as the

basis for the plant safety analyses. The

potential failure modes for the components

within the scope of this request were

evaluated for impact on instrument response

time. This evaluation confirmed that the

remaining TS-required testing is sufficient to

identify failure modes or degradations in

instrument response times and to ensure that

operation of the instrumentation within the

scope of this request is within acceptable

limits. As a result, it has been concluded that

plant and system response to an initiating

event will remain in compliance with the

assumptions of the safety analysis.

Further, although not explicitly evaluated,

the proposed changes will provide an

improvement to plant safety and operation by

reducing the time safety systems are

unavailable, reducing the potential for safety

system actuations, reducing plant shutdown

risk, limiting radiation exposure to plant

personnel, and eliminating the diversion of

key personnel resources to conduct

unnecessary testing. Therefore, IP has

concluded that this request will result in an

overall increase 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: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Attorney for licensee: Sheldon Zabel,

Esq., Schiff, Hardin and Waite, 7200

Sears Tower, 233 Wacker Drive,

Chicago, Illinois 60606

NRC Project Director: Gail H. Marcus

Northern States Power Company,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request:

December 11, 1995

Description of amendment request:

The proposed amendment would

modify Technical Specification (TS)

Section 4.7, Surveillance Requirements

for Primary Containment Automatic

Isolation Valves. Specifically, the

proposed amendment would delete TS

Surveillance Requirement 4.7.D.4,

which requires replacement of the seat

seals for the drywell and suppression

chamber purge and vent valves every 5

years.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed amendment will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

An evaluation of the operational

performance of the 18-inch purge and vent

valves has concluded that deletion of the

Monticello Technical Specification

surveillance requirement 4.7.D.4 will have no

adverse impact on the seat leakage

performance of these primary containment

isolation valves, no adverse impact on the

testing performed in accordance with 10 CFR

50, Appendix J, and thus no adverse impact

on the containment isolation function of

these primary containment isolation valves.

The material of which the T-shaped

elastomer seat is comprised of has been

found to withstand normal and accident

thermal exposures for the design life of the

plant based on a thermal aging analysis.

Radiation effects will not have an adverse

impact on the elastomer seat material.

Therefore, this amendment will not cause a

significant increase in the probability or

consequences of an accident previously

evaluated for the Monticello plant.

The proposed amendment will not create

the possibility of a new or different kind of

accident from any accident previously

analyzed.

The proposed change to the Technical

Specifications for the Primary Containment

Purge and Vent valves does not alter the

function of these components or their

interrelationships with other systems.

Therefore, this amendment will not create

the possibility of a new or different kind of

accident from any accident previously

analyzed.

The proposed amendment will not involve

a significant reduction in the margin of

safety.

The operating experience of these valves

has demonstrated that the testing performed
3502 Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices

in accordance with 10 CFR 50, Appendix J,

provides a high level of confidence in the

ability of these valves to perform their safety

function with respect to valve leak tightness.

The proposed amendment 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: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037

NRC Project Director: John N. Hannon

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests:

December 27, 1995

Description of amendment requests:

The amendments would revise the

combined Technical Specifications (TS)

3/4.6.1.1, ‘‘Containment Integrity;’’ 3/

4.6.1.2, ‘‘Containment Leakage;’’ 3/

4.6.1.3, ‘‘Containment Air Locks;’’ 3/

4.6.1.6, ‘‘Containment Structural

Integrity;’’ 3/4.6.3, ‘‘Containment

Isolation Valves;’’ and their associated

Bases; and would add TS 6.8.4.j,

‘‘Containment Leakage Rate Testing

Program,’’ to implement the

performance-based leakage rate testing

program, as permitted by 10 CFR Part

50, Appendix J. These changes will

support the implementation of the

performance-based testing of Option B

to Appendix J for Types A, B, and C

containment leakage rate testing and the

appropriate rescheduling of testing. The

amendment changes the TS to

implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leakage Test Program.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to Technical

Specification (TS) 3/4.6.1.1, 3/4.6.1.2, 3/

4.6.1.3, 3/4.6.1.6, 3/4.6.3, and the addition of

6.8.4 j., to implement the performance-based

Containment Leakage Rate Testing Program

have no effect on plant operation. The

proposed changes only provide mechanisms

within the TS for implementing a

performance-based methodology for

determining the frequency of leak rate testing

that has been approved by the Commission.

The test type and test method used for testing

would not be changed. The test acceptance

criteria would not be changed, and

containment leakage will continue to be

maintained within the required limits.

Directly referencing the Containment

Leakage Rate Testing Program for

containment ILRT [integrated leak rate

testing] and LLRT [local leak rate test]

requirements does not involve any

modification to plant equipment or affect the

operation or design basis of the containment.

Leakage rate testing is not a precursor to or

an initiating event for any accident.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes only allow for the

implementation of 10 CFR 50, Appendix J,

Option B, testing frequencies and do not

involve any modifications to any plant

equipment or affect the operation or design

basis of the containment. The proposed

changes do not affect the response of the

containment during a design basis accident.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes do not affect or

change a Safety Limit or affect plant

operations. The changes only implement the

allowed 10 CFR 50, Appendix J, Option B

testing frequencies that have been

determined by the Commission not to

involve a safety concern. The testing method,

acceptance criteria, and basis for testing are

not changed and still provide assurance that

the containment will provide its intended

function.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman

Saxton Nuclear Experimental

Corporation (SNEC), Docket No. 50–

146, Saxton Nuclear Experimental

Facility (SNEF), Bedford County,

Pennsylvania

Date of amendment request:

November 21, 1995.

Description of amendment request:

The proposed amendment would

change the license and technical

specifications to add GPU Nuclear

Corporation (GPUN) as a licensee for the

SNEF along with SNEC and would

transfer from SNEC to GPUN all

management-related responsibilities for

the SNEF. Responsibility for safely

maintaining the containment vessel and

performing characterization activities

would change from SNEC to GPUN.

Technical specification organizational

positions would be changed from SNEC

titles to GPUN titles. GPUN would take

responsibility from SNEC for

administration of all SNEF functions,

for radiation safety activities, and for

providing on-site management and

continuing oversight of production

activities. The appointment of members

to the Saxton Radiation Safety

Committee and the reporting of the

Committee would change from the

SNEC President to the GPUN Vice

President of the Nuclear Services

Division. The GPUN President would

have the authority to request audits and

would receive audit reports instead of

the SNEC President. Procedure control

methodology and the administrative

procedure for procedures would be

changed from SNEC procedures to

GPUN procedures. The responsibility

for records retention and reporting

would change from SNEC to GPUN. The

organization chart for the facility would

be changed to reflect the addition of

GPUN as a licensee.

Basis for proposed no significant

Hazards Consideration Determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes 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.

Because the proposed changes are

administrative in nature they would have no

effect on the likelihood or impact on the

potential accidents of fire, flood or

radiological hazard.
Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices 3503

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Because the proposed changes are

administrative in nature they would not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

3. Involve a significant reduction in a

margin of safety.

Because the proposed changes are

administrative in nature they would not

involve any reduction in a margin of safety.

The NRC staff has reviewed the

analysis of the licensee and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Saxton Community Library,

911 Church Street, Saxton,

Pennsylvania 16678 Attorney for the

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

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Seymour H.

Weiss

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: January

4, 1996 (TS 95–22)

Description of amendment request:

The proposed change would extend the

functional testing interval for the

following isolation radiation monitor

instruments from monthly to quarterly:

(1) Engineered Safety Feature Actuation

System Instrumentation Surveillance

Requirements Table 4.3–2, Item 3.c.3,

Containment Purge Air Exhaust Monitor

Radioactivity-High; (2) Radiation

Monitoring Instrumentation

Surveillance Requirements Table 4.3–3,

Item 1.a, Fuel Storage Pool Area

Radiation Monitor; (3) Table 4.3–3, Item

2.a, Containment Purge Air Exhaust; (4)

Table 4.3–3, Item 2.b.i, Containment

Gaseous Activity RCS Leakage

Detection; (5) Table 4.3–3, Item 2.b.ii,

Containment Particulate Activity RCS

Leakage Detection; and (6) Table 4.3–3,

Item 2.c, Control Room Isolation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Review of the past history for the affected

and similar radiation monitors revealed that

extending the functional testing interval for

these monitors will not adversely affect

system operability and will effectively

increase system availability. These radiation

monitors are not accident initiating

equipment, thus increasing the surveillance

interval on these monitors will not affect the

probability of any accident previously

evaluated. Based on the above statements, it

is concluded that the probability or

consequences of an accident previously

evaluated is not increased.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

No new type of accident or malfunction

will be created since the radiation monitors

are not accident initiating equipment. The

proposed change merely increases the

functional testing interval for the affected

radiation monitors, and does not change the

method and manner of plant operation. The

safety design bases in the Updated Final

Safety Analysis Report have not been altered.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously analyzed.

3. Involve a significant reduction in a

margin of safety.

The proposed changes do not change the

plant configuration in a way that introduces

a new potential hazard to the plant and do

not involve a significant reduction in the

margin of safety. The proposed changes do

not affect applicable safety analysis

acceptance criteria and will not affect system

operating conditions. Additionally, plant

operating experience with similar monitors

has shown that there has not been additional

failures due to the quarterly testing

frequency. Thus, it is concluded that the

margin of safety is not reduced.

The NRC 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: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

November 22, 1995

Description of amendment request:

The proposed amendment replaces the

requirements associated with the boron

dilution mitigation system (BDMS) in

the Wolf Creek Generating Station

Technical Specifications with alarms,

indicators, procedures, and controls to

assure proper resolution of potential

inadvertent boron dilution events.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

. The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The only event potentially impacted by the

proposed change is the inadvertent boron

dilution event. The discussion of the

probability and consequences of an

inadvertent boron dilution event at WCGS is

provided in USAR [Updated Safety Analysis

Report] Section 15.4.6. Primarily, the

proposed changes revise the method of

detecting and mitigating the event. The only

aspect of the changes that impact[s] the

potential causes of an inadvertent boron

dilution event is the increased requirement to

isolate potential dilution sources in Modes 3,

4 and 5. As a result, the overall probability

of the event is slightly decreased.

The alternate methods to detect and

mitigate this event achieve the same basic

goal as the current BDMS; to prevent a return

to critical during an inadvertent dilution

event. The proposed changes to the BDMS

will result in an improved system that will

provide an improved response to the

inadvertent boron dilution event, and that

will prevent a return to critical. Thus, it can

be concluded that the proposed change will

not significantly increase the consequences

of a postulated inadvertent boron dilution

event.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The revisions to plant procedural

requirements to either operate a reactor

coolant pump or to isolate/control potential

dilution sources does [sic] not create the

potential for a new or different kind of

accident because these new requirements are

configurations which have always been

allowed. Similarly, the new normal position

for the letdown divert valve does not create

a new or different accident because the new

normal position has always been an allowed

position. The other procedural changes only

increase the plant operators’ awareness of

potential boron dilution problems or provide

the steps needed to respond to available

indications and alarms to mitigate the

potential event. As a result, these procedural

changes do not create the possibility of a new

or different kind of accident.

The proposed changes also include

addition of new redundant VCT high level

alarms and a new alarm indicating that the
3504 Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices

letdown divert valve is not in the ‘‘VCT’’

position. Because the alarms are passive, they

do not create the possibility of a new or

different kind of accident.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The design criterion and margin of safety

for the current BDMS is that the dilution

event is terminated prior to the loss of all

shutdown margin. The same criterion will be

met following the implementation of the

proposed changes. Therefore, there is no

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 20, 1995

Description of amendment request:

This amendment request proposes to

revise Technical Specification 3/4.6.1.1,

‘‘Containment Integrity,’’ and 3/4.6.1.3,

‘‘Containment Air Locks,’’ and to add

Technical Specification 6.8.4i,

‘‘Containment Leakage Rate Testing

Program,’’ to implement the new

performance-based leakage rate testing

program as permitted by 10 CFR 50,

Appendix J. Also, Technical

Specification 1.7e, ‘‘Containment

Integrity,’’ would be revised to reference

Technical Specification 4.6.1.1.c. These

proposed changes will implement the

performance-based testing of Option B

to Appendix J, for Type A, B, and C

containment leak testing by referring to

Regulatory Guide 1.163, ‘‘PerformanceBased Containment Leakage-Test

Program.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to Technical

Specifications 3/4.6.1.1 and 3/4.6.1.3, and

the addition of Technical Specification 6.8.4i

to implement the new performance based

Containment Leakage Rate Testing Program,

have no effect on plant operation. The

proposed changes only provide mechanisms

within the technical specifications for

implementing a performance-based

methodology, for determining the frequency

of leak rate testing, which has been approved

by the NRC. The test type and test method

used for testing would not be changed. The

test acceptance criteria would not be

changed, and containment leakage will

continue to be maintained within the

required limits.

Directly referencing the Containment

Leakage Rate Testing Program for

containment integrated leak rate test and

local leak rate test requirements does not

involve any modification to plant equipment

or affect the operation or design basis of the

containment. Leakage rate testing is not a

precursor to or an initiating event for any

accident.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes only allow for

implementation of 10 CFR 50, Appendix J,

Option B, testing frequencies and do not

involve any modifications to any plant

equipment or affect the operation or design

basis of the containment. The proposed

changes do not affect the response of the

containment during a design basis accident.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes do not affect or

change a Safety Limit, any limiting condition

for operation or affect plant operations. The

changes only implement the allowed Option

B testing frequencies that have been

determined by the NRC not to involve a

safety concern. The testing method,

acceptance criteria, and bases are not

changed and still provide assurance that the

containment will provide its intended

function.

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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.
Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices 3505

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:

November 14, 1995, as supplemented

January 4, 1996.

Brief description of amendments: The

amendments revise the Technical

Specifications to incorporate 10 CFR

Part 50, Appendix J, ‘‘Primary Reactor

Containment Leakage Testing for WaterCooled Power Reactors,’’ Option B.

Technical Specification changes for the

LaSalle facility will be addressed under

separate correspondence.

Date of issuance: January 11, 1996

Effective date: January 11, 1996

Amendment Nos.: 148, 142, 169, and

165

Facility Operating License Nos. DPR–

19, DPR–25, DPR–29 and DPR–30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 7, 1995 (60 FR

62896). The January 4, 1996,

supplement provided a specific

implementation date for the requested

amenement. This information was

within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated January 11, 1996. 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.

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of application for amendment:

November 14, 1995

Brief description of amendment: The

amendment revises the Haddam Neck

Technical Specifications (TS) to provide

an one-time exception to TS 3.9.12, ’’

Fuel Building Storage Air Cleanup

System,’’ to allow the fuel storage

building air cleanup system to be

inoperable for a limited duration during

intervals in which new fuel rack

modules will be moved into and old

fuel rack modules will be moved out of

the fuel storage building.

Date of Issuance: January 17, 1996

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 187

Facility Operating License No. DPR–

61. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 28, 1995 (60 FR

58688) The Commission’s related

evaluation of this amendment is

contained in a Safety Evaluation dated

January 17, 1996 No significant hazards

consideration comments received: No.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Consumers Power Company, Docket

No. 50–155, Big Rock Point Plant,

Charlevoix County, Michigan Date of

application for amendment: November

8, 1995, as supplemented November 17,

1995

Brief description of amendment: The

amendment removes the prescriptive

Type A containment leakage test rate

frequency of 40 plus or minus 10

months and adds a reference to perform

containment leakage rate tests in

accordance with the criteria specified in

Appendix J of 10 CFR Part 50 as

modified by approved exemptions. In

addition, the amendment revises the test

pressure for Type B and C testing to

correct a typographical error.

Date of issuance: January 16, 1996

Effective date: January 16, 1996

Amendment No.: 117

Facility Operating License No. DPR–6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62489) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 16, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770.

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:

March 29, 1995, as supplemented by

letters dated September 18 and

November 16, 1995

Brief description of amendments: The

amendments revise Technical

Specification requirements for the Low

Temperature Overpressure Protection

system and update the heatup and

cooldown curves for both units.

Date of issuance: January 11, 1996

Effective date: As of the date of

issuance to be implemented within 60

days

Amendment Nos.: Unit 1 - 162; Unit

2 - 144

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications. Date of initial

notice in Federal Register: September

27, 1995 (60 FR 49933) The September

18 and November 16, 1995, letters

provided clarifying information that did

not change the scope of the March 29,

1995, 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 January 11, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

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:

December 2, 1994

Brief description of amendments: The

amendments replace Appendix B,

‘‘Environmental Technical

Specifications,’’ with an Environmental

Protection Plan (Nonradiological) and

revise the Operating Licenses to reflect

these changes.

Date of issuance: December 19, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: Unit 1 - 199 - Unit

2 - 140

Facility Operating License Nos. DPR–

57 and NPF–5. Amendments revised the

Technical Specifications and Operating

Licenses.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 502)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated December 19, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513
3506 Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

March 17, 1995, as supplemented by

letter dated July 6, 1995

Brief description of amendments: The

amendments revise Technical

Specification 3/4.9.4, Containment

Building Penetrations, to allow the

personnel airlock to be open during core

alterations or movement of irradiated

fuel within the containment.

Date of issuance: November 30, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 92 and 70

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35077) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated November 30, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50–458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: August

17, 1995, as supplemented by letters

dated November 22, and December 18,

20, and 27, 1995

Brief description of amendment: The

amendment revised the primary

containment air lock technical

specifications to allow the air locks to

be open in Mode 5 (refueling) during

core alterations except for movement of

recently irradiated fuel. All other

provisions of the August 17, 1995,

requests are defered.

Date of issuance: January 11, 1996

Effective date: January 11, 1996

Amendment No.: 85

Facility Operating License No. NPF–

47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 13, 1995 (60 FR

47619) The additional information

contained in the supplemental letters

dated November 22, and December 18,

20, and 27, 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 Janaury 11, 1996. No

significant hazards consideration

comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50–458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request:

November 20, 1995

Brief description of amendment: The

proposed amendment revised the

technical specifications to eliminate the

response time testing requirements for

selected Reactor Protection System

Instrumentation.

Date of issuance: January 11, 1996

Effective date: January 11, 1996

Amendment No.: 86

Facility Operating License No. NPF–

47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62492) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

Janaury 11, 1996. No significant hazards

consideration comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

May 25, 1995 (AEP:NRC:1071T)

Brief description of amendments: The

amendments incorporate a cycle- and

burnup-dependent peaking factor

penalty in the Core Operating Limits

Report and add an appropriate reference

to the COLR and update the topical

report reference in the Technical

Specifications.

Date of issuance: January 4, 1996

Effective date: January 4, 1996, with

full implementation within 45 days

Amendment Nos.: Unit 1, 206, Unit 2,

190

Facility Operating License Nos. DPR–

58 and DPR–74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 4, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Northeast Nuclear Energy Company,

Docket No. 50–245, Millstone Nuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

October 3, 1995

Brief description of amendment: The

amendment removes the Limiting

Condition for Operation (LCO) and

Surveillance Requirements for the lossof-normal power (LNP) trip function

from Tables 3.2.2 and 4.2.1 and inserts

new LCO 3.2.F and Surveillance

Requirement 4.2.F. In addition, the

amendment adds a new table to specify

the required LNP instrumentation for

each bus, updates the Table of Contents,

makes some editorial changes, and

revises the associated Bases section.

Date of issuance: January 17, 1996

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 92

Facility Operating License No. DPR–

21. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 4, 1995 (60 FR

62111) The Commission’s related

evaluation of the amendment is

contained in a Safety evaluation dated

January 17, 1996. 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. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

September 11, 1995, as supplemented

November 15, 1995.

Brief description of amendment: The

amendment changes Technical

Specification Sections 3.4.8 and 3.9.9,

Tables 2.2–1, 3.3–3, 3.3–5 and 3.3–8,

and Bases Sections 3/4.2.1, 3/4.4.8 and

3/4.11.2.1. These changes combine

several different administrative changes

which will correct typographical errors,

provide clarifications, or make editorial

changes.

Date of issuance: January 17, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 194
Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices 3507

Facility Operating License No. DPR–

65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52933) The November 15, 1995, letter

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

January 17, 1996. 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. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

September 29, 1995, as supplemented

November 9, 1995.

Brief description of amendment: The

amendment provides three changes to

the Technical Specifications (TS)

relating to the pressurizer safety valves

(PSV) and the main steam safety valves

(MSSV).

The first change is to TS 3.4.2.1 and

3.4.2.2 and involves relaxing the asfound setpoint tolerance for the

pressurizer safety valves (PSVs) and the

main steam safety valves (MSSVs) from

the current value of plus or minus 1%

to plus or minus 3%. Table 4.7–1 is also

modified to correct the as-found

tolerance for the MSSV from plus or

minus 1% to plus or minus 3%. Notes

are added to TS 3.4.2.2 and Table 4.7–

1 which specify that the lift setting

should be determined at nominal

operating conditions and should be set

at plus or minus 1% of the lift setting.

For the second change, Surveillance

Requirement 4.7.1.1 and Table 4.7–1 are

modified to eliminate the need to verify

the orifice size of each MSSV.

The third change modifies the

statement for TS 3.7.1.1 so that if a

MSSV is inoperable and compensating

action cannot be taken, the plant must

be brought to hot shutdown (Mode 4)

within 12 hours instead of cold

shutdown (Mode 5) in 30 hours.

Date of issuance: January 18, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 195

Facility Operating License No. DPR–

65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 25, 1995 (60 FR

54723) The November 9, 1995, letter

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

January 18, 1996. 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 30, 1995

Brief description of amendments: The

amendments change the technical

specification requirements for control

rod drive scram accumulator and

charging water header minimum

pressure.

Date of issuance: January 11, 1996

Effective date: Unit 2, as of date of

issuance, to be implemented

concurrently with Amendment 210,

issued August 30, 1995; Unit 3, as of

date of issuance, to be implemented

concurrently with Amendment 214,

issued August 30, 1995.

Amendments Nos.: 211 and 216

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 8, 1995 (60 FR

63073) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 11, 1996 No significant hazards

consideration comments received: No

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

PECO Energy Company, Public Service

Electric and Gas 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 21, 1995

Brief description of amendments: The

amendments change the test pressure

requirements for the high pressure

coolant injection system and the reactor

core isolation cooling system

surveillance tests. The amendments also

change Section 5.5.7 of the technical

specifications to eliminate reference to a

section which was previously

eliminated.

Date of issuance: January 11, 1996

Effective date: As of date of issuance.

Amendments Nos.: 212 and 217

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 5, 1995 (60 FR

62271) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 11, 1996 No significant hazards

consideration comments received: No

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

PECO Energy Company, Public Service

Electric and Gas 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:

December 19, 1995

Brief description of amendments:

These amendments change the

ventilation filter test program bypass

and penetration leakage test acceptance

criteria from less than 0.05 percent to

less than 1.0 percent. The change

corrects an administrative error that

occurred during the development of the

Peach Bottom Improved Technical

Specifications which were issued as

Amendments 210 and 214 to the Peach

Bottom licenses on August 30, 1995.

Date of issuance: January 16, 1996

Effective date: Unit 2, effective as of

date of issuance, to be implemented

concurrently with Amendment 210,

issued August 30, 1995; Unit 3, effective

as of date of issuance, to be

implemented concurrently with

Amendment 214, issued August 30,

1995.

Amendments Nos.: 213 and 218

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (60 FR 66997,

December 27, 1995). That notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards
3508 Federal Register / Vol. 61, No. 21 / Wednesday, January 31, 1996 / Notices

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by January 26, 1996,

but indicated that if the Commission

makes a final no significant hazards

consideration determination any such

hearing would take place after issuance

of the amendment. The Commission’s

related evaluation of the amendments,

finding of exigent circumstances, and

final determination of no significant

hazards consideration are contained in

a Safety Evaluation dated January 16,

1996

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.

Dated at Rockville, Maryland, this 23rd day

of January 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation.

[Doc. 96–1683 Filed 1–30–96; 8:45 am]

BILLING CODE 7590–01–F

Draft Regulatory Guide; Issuance,

Availability

The Nuclear Regulatory Commission

has issued for public comment a draft of

a guide planned for its Regulatory Guide

Series. This series has been developed

to describe and make available to the

public such information as methods

acceptable to the NRC staff for

implementing specific parts of the

Commission’s regulations, techniques

used by the staff in evaluating specific

problems or postulated accidents, and

data needed by the staff in its review of

applications for permits and licenses.

The draft guide is a proposed

Revision 1 to Regulatory Guide 5.15,

and it is temporarily identified as DG–

5005, ‘‘Tamper-Indicating Seals for the

Protection and Control of Special

Nuclear Material.’’ The guide will be in

Division 5, ‘‘Materials and Plant

Protection.’’ This regulatory guide is

being revised to describe features of

security seal systems and types of seals

that are acceptable to the NRC staff for

tamper-safing containers of special

nuclear material.

This draft guide is being issued to

involve the public in the early stages of

the development of a regulatory position

in this area. It has not received complete

staff review and does not represent an

official NRC staff position.

Public comments are being solicited

on the guide. Comments should be

accompanied by supporting data.

Written comments may be submitted 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.

Copies of comments received may be

examined at the NRC Public Document

Room, 2120 L Street NW., Washington,

DC. Comments will be most helpful if

received by April 12, 1996.

Although a time limit is given for

comments on this draft guide,

comments and suggestions in

connection with items for inclusion in

guides currently being developed or

improvements in all published guides

are encouraged at any time.

Comments may be submitted

electronically, in either ASCII text or

Wordperfect format (version 5.1 or

later), by calling the NRC Electronic

Bulletin Board on FedWorld. The

bulletin board may be accessed using a

personal computer, a modem, and one

of the commonly available

communications software packages, or

directly via Internet.

If using a personal computer and

modem, the NRC subsystem on

FedWorld can be accessed directly by

dialing the toll free number: 1–800–

303–9672. Communication software

parameters should be set as follows:

parity to none, data bits to 8, and stop

bits to 1 (N,8,1). Using ANSI or VT–100

terminal emulation, the NRC NUREGs

and RegGuides for Comment subsystem

can then be accessed by selecting the

‘‘Rules Menu’’ option from the ‘‘NRC

Main Menu.’’ For further information

about options available for NRC at

FedWorld, consult the ‘‘Help/

Information Center’’ from the ‘‘NRC

Main Menu.’’ Users will find the

‘‘FedWorld Online User’s Guides’’

particularly helpful. Many NRC

subsystems and data bases also have a

‘‘Help/Information Center’’ option that

is tailored to the particular subsystem.

The NRC subsystem on FedWorld can

also be accessed by a direct dial phone

number for the main FedWorld BBS,

703–321–3339, or by using Telnet via

Internet, fedworld.gov. If using 703–

321–3339 to contact FedWorld, the NRC

subsystem will be accessed from the

main FedWorld menu by selecting the

‘‘Regulatory, Government

Administration and State Systems,’’

then selecting ‘‘Regulatory Information

Mall.’’ At that point, a menu will be

displayed that has an option ‘‘U.S.

Nuclear Regulatory Commission’’ that

will take you to the NRC Online main

menu. The NRC Online area also can be

accessed directly by typing ‘‘/go nrc’’ at

a FedWorld command line. If you access

NRC from FedWorld’s main menu, you

may return to FedWorld by selecting the

‘‘Return to FedWorld’’ option from the

NRC Online Main Menu. However, if

you access NRC at FedWorld by using

NRC’s toll-free number, you will have

full access to all NRC systems but you

will not have access to the main

FedWorld system.

If you contact FedWorld using Telnet,

you will see the NRC area and menus,

including the Rules menu. Although

you will be able to download

documents and leave messages, you will

not be able to write comments or upload

files (comments). If you contact

FedWorld using FTP, all files can be

accessed and downloaded but uploads

are not allowed; all you will see is a list

of files without descriptions (normal

Gopher look). An index file listing all

files within a subdirectory, with

descriptions, is included. There is a 15-

minute time limit for FTP access.

Although FedWorld can be accessed

through the World Wide Web, like FTP

that mode only provides access for

downloading files and does not display

the NRC Rules menu.

For more information on NRC bulletin

boards call Mr. Arthur Davis, Systems

Integration and Development Branch,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555, telephone

(301)415–5780; e-mail AXD3@nrc.gov.

For more information on this draft

regulatory guide, contact S.D. Frattali at

the NRC, telephone (301)415–6261; email SDF@nrc.gov.

Regulatory guides are available for

inspection at the Commission’s Public

Document Room, 2120 L Street NW.,

Washington, DC. Requests for single

copies of draft or final guides (which

may be reproduced) or for placement on

an automatic distribution list for single

copies of future draft guides in specific

divisions should be made in writing to

the U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

Attention: Distribution and Mail

Services Section; or by fax at (301)415–

2260. Telephone requests cannot be

accommodated. Regulatory guides are

not copyrighted, and Commission

approval is not required to reproduce

them.

(5 U.S.C. 552(a))

Dated at Rockville, Maryland, this 18th day

of January 1996.

For the Nuclear Regulatory Commission.

Frank A. Costanzi,

Deputy Director, Division of Regulatory

Applications, Office of Nuclear Regulatory

Research.

[FR Doc. 96–1878 Filed 1–30–96; 8:45 am]

BILLING CODE 7590–01–P