Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18621

whether to retain or sell the Plan’s

Interest.

The Department has considered the

entire record, including the written

comment submitted by the applicant,

and has determined to grant the

exemption as it was proposed.

FOR FURTHER INFORMATION CONTACT: Gary

H. Lefkowitz of the Department,

telephone (202) 219–8881. (This is not

a toll-free number.)

General Information

The attention of interested persons is

directed to the following:

(1) The fact that a transaction is the

subject of an exemption under section

408(a) of the Act and/or section

4975(c)(2) of the Code does not relieve

a fiduciary or other party in interest or

disqualified person from certain other

provisions to which the exemptions

does not apply and the general fiduciary

responsibility provisions of section 404

of the Act, which among other things

require a fiduciary to discharge his

duties respecting the plan solely in the

interest of the participants and

beneficiaries of the plan and in a

prudent fashion in accordance with

section 404(a)(1)(B) of the Act; nor does

it affect the requirement of section

401(a) of the Code that the plan must

operate for the exclusive benefit of the

employees of the employer maintaining

the plan and their beneficiaries;

(2) These exemptions are

supplemental to and not in derogation

of, any other provisions of the Act and/

or the Code, including statutory or

administrative exemptions and

transactional rules. Furthermore, the

fact that a transaction is subject to an

administrative or statutory exemption is

not dispositive of whether the

transaction is in fact a prohibited

transaction; and

(3) The availability of these

exemptions is subject to the express

condition that the material facts and

representations contained in each

application accurately describes all

material terms of the transaction which

is the subject of the exemption.

Signed at Washington, D.C., this 6th day of

April, 1995.

Ivan Strasfeld,

Director of Exemption Determinations,

Pension and Welfare Benefits Administration,

Department of Labor.

[FR Doc. 95–8915 Filed 4–11–95; 8:45 am]

BILLING CODE 4510–29–P

NUCLEAR REGULATORY

COMMISSION

Documents Containing Reporting or

Recordkeeping Requirements; Office

of Management and Budget (OMB)

Review

AGENCY: Nuclear Regulatory

Commission (NRC).

ACTION: Notice of the OMB review of

information collection.

SUMMARY: The Nuclear Regulatory

Commission has recently submitted to

OMB for review the following proposal

for collection of information under the

provisions of the Paperwork Reduction

Act of 1980 (44 U.S.C. Chapter 35).

1. Type of submission, new, revised,

or extension: Revision.

2. The title of the information

collection: Proposed Rules, 10 CFR Part

52, Appendix A, ‘‘Design Certification

Rule for the U.S. Advanced Boiling

Water Reactor,’’ and Appendix B,

‘‘Design Certification Rule for the

System 80+ Design.’’

13. The form number if applicable:

Not applicable.

4. How often is the collection

required: Quarterly until the applicant

or licensee receives either an operating

license under 10 CFR 50, or the

Commission makes its findings under

10 CFR 52.103.

5. Who will be required or asked to

report: Applicant and holders of

construction permits and combined

licenses.

6. An estimate of the number of

annual respondents: None anticipated

in the next three years.

7. An estimate of the number of hours

needed annually to complete the

requirement or request: For both

Appendix A and B, 0 burden hours are

anticipated over the next three years.

However, when utilized, 8 hours per

respondent for reporting will be

required.

8. An indication of whether Section

3504(h), Pub. L. 96–511 applies:

Applicable.

9. Abstract: The standard design

certification rule (10 CFR 52) was

codified to establish procedures,

standards and criteria governing

standard design certification, including

informal submittal and recordkeeping

requirements. Appendices A–L to Part

52 are reserved to constitute the

standard design certification for

evolutionary and passive light water

reactor design. These proposed rules

will certify the Advanced Boiling Water

Reactor (ABWR) and System 80+

Standard designs, will be mandatory for

those applicants proposing to use

Appendix A or B, and will ensure the

safety of the public.

Copies of the submittal may be

inspected or obtained for a fee from the

NRC Public Document Room, 2120 L

Street NW. (Lower Level), Washington,

DC.

Comments and questions can be

directed by mail to the OMB reviewer:

Troy Hillier, Office of Information and

Regulatory Affairs (3150–0151), NEOB–

10202, Office of Management and

Budget, Washington, DC 20503.

Comments may also be communicated

by telephone at (202) 395–3084.

The NRC Clearance Office, is Brenda

Jo Shelton, (301) 415–7233.

Dated at Rockville, Maryland, this 6th day

of April, 1995.

For the Nuclear Regulatory Commission.

Gerald F. Cranford,

Designated Senior Official for Information

Resources Management.

[FR Doc. 95–8971 Filed 4–11–95; 8:45 am]

BILLING CODE 7590–01–M

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from March 17,

1995, through March 31, 1995. The last

biweekly notice was published on

March 29, 1995 (60 FR 16181).

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
18622 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

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 May 12, 1995, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

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
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18623

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, IllinoisDocket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of amendment request: March

23, 1994, as supplemented on July 26,

1994, February 15, 1995, and February

28, 1995.

Description of amendment request: In

the submittals of March 23 and July 26,

1994, the licensee requested revisions to

the plants’ technical specifications (TSs)

to permit the use of a slightly positive

reactor core moderator temperature

coefficient (MTC). The February 15,

1995, submittal requested approval to

expand the operating limits report

(OLR) to include a cycle specific MTC

value and requested approval to

maintain the MTC value within the

limits specified in the OLR. The

maximum upper MTC limit would be

specified in the TSs. The February 28,

1995, submittal provided a revised

Significant Hazards Consideration. This

supplements the information that was

published in the Federal Register on

August 31, 1994 (59 FR 45037).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

An analysis program was pursued by

Commonwealth Edison to justify a positive

MTC, reduced reactor coolant system thermal

design flow, and increased steam generator

tube plugging levels. This analysis identified

a need for corresponding increases in the

boron concentration levels in the refueling

water storage tank (RWST) and safety

injection accumulators to assure

subcriticality requirements are met following

a postulated loss-of-coolant accident (LOCA).

The increases in boron concentration are

based on the maximum upper limit of the

MTC. The corresponding Technical

Specification changes required as a result of

this analysis program were previously

approved by the NRC, including the

increases in boron concentration limits, with

the exception of the positive MTC change.

The safety analyses necessary to support this

program are documented in WCAP-13964.

The results were reviewed by

Commonwealth Edison and found to be

acceptable. All Departure from Nucleate

Boiling Ratio (DNBR) design limits were

determined such that there was a 95 percent

probability at a 95 percent confidence level

that DNB would not occur on the most

limiting fuel rod for any Condition I or

Condition II event. The present Technical

Specification limit for Nuclear Enthalpy Rise

Hot Channel Factor, ... , of less than 1.65

ensures that the DNB design basis stated

above would be met, thus fuel integrity will

not be challenged.

The accidents which are sensitive to MTC

were analyzed as part of the overall program

and the results were found to be acceptable.

The safety functions of the evaluated systems

and components remain unchanged. The

analysis performed using the increased MTC

value does not affect the integrity of the

safety related systems and components such

that their function to control radiological

consequences is affected and all fission

barriers will remain intact. The effects on

offsite doses have been considered. The

incorporation of a positive MTC, in

conjunction with the previously approved

reduction in reactor coolant system thermal

design flow rate and increase in steam

generator tube plugging levels, will result in

a small increase in offsite doses; however, the

total doses remain a small fraction of the 10

CFR 100 limits. As such, the accident

analysis acceptance criteria continue to be

satisfied.

On a cycle-by-cycle basis, a deterministic

evaluation of the impact on ATWS risk will

be performed. An Unfavorable Exposure

Time (UET) will be calculated, where UET is

defined as the amount of time during the

operating cycle for which the reactivity

feedback is not sufficient to prevent Reactor

Coolant System (RCS) pressure from

exceeding 3200 psig for a given plant

configuration. The UET methodology is

consistent with the Westinghouse Owner’s

Group methodology presented in WCAP

11992, ‘‘ATWS Rule Administration Process’’

and WCAP 11993, ‘‘Assessment of

Compliance with ATWS Rule Basis for

Westinghouse PWRs’’. Corrective actions will

be taken, as necessary, to assure a UET of less

than 5 percent of cycle length.

The relocation of the cycle-specific core

operating limits for the MTC from the

Technical Specifications has no influence or

impact on the probability or consequences of

any accident previously evaluated. Byron

and Braidwood Stations will continue to

operate within the cycle-specific MTC limits

contained in the OLR. The proposed

amendment will require exactly the same

action to be taken when the OLR limits are

exceeded as are required by the current

Technical Specification. Any change to the

MTC values in the OLR will be performed

based on NRC-approved methodology as

delineated in Section 6.9.1.9 of the Technical

Specifications. Each accident analysis

addressed in the Updated Final Safety

Analysis Report (UFSAR) will be examined

with respect to changes in cycle dependent

parameters, which are obtained from

application of NRC-approved reload design

methodologies, to ensure that the transient

evaluation of new reloads are bounded by

previously accepted analysis. This

examination, which will be performed under

the requirements of 10 CFR 50.59, ensures

that future reloads will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Therefore, implementation of a positive

MTC will not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed changes do not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

The methodology and manner of plant

operation as a result of the proposed changes

is unaffected. Implementation of a positive

MTC does not impact the safe operation of

the reactor provided that the Limiting

Conditions for Operation (LCOs) and the

associated action requirements are satisfied.

The assumptions do not create any new

failure modes that could adversely impact

safety related equipment. The reload safety

limits and LCOs in the plant Technical

Specifications will be evaluated and satisfied

for each future reload core design via the 10

CFR 50.59 process. All DNBR limits have

been satisfied. Currently installed equipment

will not be operated in a manner different

than previously designed. No new credible

limiting single failure has been created. No

new or different accidents or failure modes

have been identified for any systems or

components important to safety.

The relocation of the cycle specific MTC

values to the OLR will not create the

possibility of a new or different type of

accident. No safety related equipment or

safety function will be altered as a result of

this proposed change. The cycle specific

values are calculated using NRC-approved

methods and submitted to the NRC to allow

the Staff to continue to trend these limits.

The Technical Specifications will continue to

require operation within the analyzed core

operating limits and appropriate actions will

be taken, when, or if, the limits are exceeded.
18624 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

Therefore, there is not a potential for

creating the possibility of a new or different

type of accident from any accident

previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The performance and integrity of the

evaluated safety related systems and

components are not affected by the proposed

change to the MTC. The radiological

consequences of all previously analyzed

accidents remain within acceptable limits.

The proposed change to the MTC will have

no effect on the availability, operability, or

performance of the evaluated safety related

systems or components. The incorporation of

a positive MTC, in conjunction with the

previously approved reduction in reactor

coolant system thermal design flow rate and

increase in steam generator tube plugging

levels, will result in a small increase in

offsite doses; however, the total doses remain

a small fraction of the 10CFR100 limits. The

methodology, discussed in Attachment E,

describes the determination and use of the

UET values in the calculation of the Primary

Pressure Relief node for the ATWS event tree

to determine an overall ATWS risk value.

The methodology will be used by ComEd to

ensure that a core designed with a positive

MTC will not result in an unacceptable risk

to core damage frequency due to an ATWS

event. The margin of safety associated with

the licensing basis safety analysis is not

significantly reduced by the proposed

changes. All acceptance criteria for the

specific UFSAR Chapter 15 safety analyses

(non-LOCA and LOCA) have been

satisfactorily evaluated and verified using

NRC approved methodologies.

The margin of safety is not affected by the

relocation of the cycle specific MTC limits

from the Technical Specifications. The

proposed amendment continues to require

operation within the core limits as

determined by the NRC-approved reload

design and safety analysis methodologies.

Appropriate actions will be taken, when, or

if, limits are exceeded.

The development of the MTC limits for

future reloads will continue to conform to

those methods described in the NRCapproved documentation. In addition, each

future reload will involve a 10 CFR 50.59

safety review to assure that operation of the

unit within the cycle specific limits will not

involve a reduction in the margin of safety

as defined in the basis for any Technical

Specification.

Therefore, there is no significant reduction

in the margin of safety as defined in the bases

of any Technical Specification.

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

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

Library, 109 N. Franklin, P.O. Box 434,

Byron, Illinois 61010; for Braidwood,

the Wilmington Township Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690

NRC Project Director: Robert A. Capra

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: October

20, 1992

Description of amendment request:

The proposed amendment would

comply with the requirements of

Amendment 135 to the Palisades

Operating License, dated February 11,

1991, which included a change to

Technical Specification 5.3.1a, Primary

Coolant System. The safety evaluation

for Amendment 135 included a

requirement that changes to Section 4.2

of the Palisades Final Safety Analysis

Report (FSAR) be made through a

formal amendment process. The

proposed FSAR change is a result of the

steam generator replacement project and

includes the following: (1) deletion of a

design load since this was not treated as

a necessary design condition in the new

steam generators; (2) a change in the

feedwater temperature from 70°F to

40°F, since this assumption was

changed in the analysis for the

replacement steam generators; and (3)

editorial changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The following summary supports the

finding that the proposed change would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The probability of an accident previously

evaluated in the FSAR will not be increased

by deleting the design load change of 15%

per minute or decreasing the minimum

feedwater temperature from 70°F to 40°F.

There is no design requirement that the plant

be capable of 15% per minute load changes.

No accident has as an initial condition a 15%

per minute load change taking place, and

since this FSAR change is the result of the

replacement steam generators design, no

accident probabilities are increased. The 40°F

feedwater temperature affects the steam

generators, but nothing else is affected in the

primary coolant system (PCS). The

replacement steam generators have been

shown by the design analysis report to be

able to withstand the same number of cycles

of the addition of 40°F water as the old steam

generators could with 70°F water.

The consequences of an accident

previously evaluated in the FSAR are not

increased by either of these two changes.

Deleting the design load rate of 15% per

minute deals with normal plant operation

and would not affect the course of a Chapter

14 event since none of the Chapter 14 events

involve power level changes with respect to

the steam generators. Also, reducing the

maximum design load change rate is a

conservative change.

Lowering the feedwater temperature could

increase the consequences of the main steam

line break (MSLB) accident by increasing the

likelihood of a return to power event caused

by increased core cooling; however, the

current FSAR analysis in Section 14.14 used

32°F as the auxiliary feedwater temperature

and thus bounds [the] 40°F [temperature].

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The possibility of a new or different type

of accident is not created by these FSAR

changes. By deleting the 15% per minute

load change rate from the FSAR, the

operation of the plant is unaffected because

the 5% per minute limit on load rate change

is more limiting. There is no license

requirement to be able to change power at

15% per minute except as described in the

proposed FSAR deletion. Furthermore, FSAR

Section 4.3.7.2 states that the pressurizer

heaters cannot be uncovered by the outward

surge of water following load increases; a

10% step increase and 15% ramp increase.

FSAR Section 1.2.4.9.a states that the nuclear

steam supply system (NSSS) is capable of a

ramp change from 15% to 100% power at 5%

per minute, and at a greater rate over smaller

load changes up to a step change of 10%.

Another consideration is that the analysis

for the original steam generators was not as

detailed or exact as the analysis for the

replacement steam generators. The thermal

analysis section of the original steam

generator design analysis report states for the

three power change cases, 5% per minute,

15% per minute and a 10% step change, that

’’... the transient thermal effects of the power

changes are small and [negligible]. The

situations of significance are due to cycling

between steady state conditions at different

power levels.’’ Thus, the rate of change was

not a consideration in the original design

analysis. The replacement steam generator

analysis calculated the transient temperature

changes with respect to time, so the rate of

change was considered. Therefore, the

replacement steam generator analysis is more

accurate, but does not consider a 15% per

minute rate change. The original steam

generators were not designed for 15% per

minute power changes but could withstand

power increases from 50% to 100% [a total

of] 15,000 times without considering the rate

of power change.

Reducing the analyzed feedwater

temperature from 70°F to 40°F does not

change the possibility of whether another

type of accident or malfunction can occur

since the steam generator is analyzed for this.

3. Involve a significant reduction in a

margin of safety.

The margin of safety as defined by plant

licensing basis is not reduced due to the

replacement steam generators not being

analyzed for a 15% per minute power ramp
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18625

because the 15% per minute ramp rate was

not a licensing basis of the plant design. The

original plant Safety Evaluation Report does

not mention the design power ramp rates.

The basis for Technical Specification 3.1.2

states that all components are designed to

withstand the effects of cyclic loads due to

primary coolant system temperature and

pressure changes induced by load changes,

trips, and start-ups and shutdowns. FSAR

Section 4.2.2 is referenced. The change of

eliminating the analyzed ability to make 15%

per minute power changes does not reduce

the margin of safety because:

a. the plant is not operated in a manner

wherein 15% per minute power increases are

made. Rapid power decreases during

emergency conditions are not covered by this

analysis since they are not controlled to 15%

per minute but should be considered

analyzed by the 500 trips or 10% step change

analysis and,

b. the original steam generator did not use

the ramp rate in the analysis and,

c. a 15% per minute power change from

50% to 100% power is a fairly benign change

for the steam generator with respect to

pressure and temperature changes as

compared to heatups and cooldowns because

the total changes are small.

The only requirement from the NRC with

respect to the number and type of loads is

contained in Section II of the NRC Standard

Review Plan (SRP) 3.9.1 which states ’’...The

section of the applicant’s SAR which pertains

to transients will be acceptable if the

transient conditions selected for equipment

fatigue evaluation are based upon a

conservative estimate of the magnitude and

frequency of the temperature and pressure

conditions resulting from those transients.’’

’’... Transients and resulting loads and load

combinations with appropriate specified

design and service limits must provide a

complete basis for design of the reactor

coolant pressure boundary for all conditions

and events expected over the service lifetime

of the plant.’’

In the intervening years between design of

the original steam generators and the

replacement steam generators, Combustion

Engineering (ABB-CE) decided that a 15%

per minute power ramp rate was beyond

what was necessary and expected to occur.

This position was acceptable to the NRC

since ABB-CE letter CPC-90-170, dated

October 24, 1990, states that the replacement

steam generators are identical in design to

the Palo Verde (Arizona Public Service)

steam generators. (The ABB-CE letter was

concerned with the stress analysis for steam

line breaks, therefore, the reference to being

identical was with respect to that stress

analysis.)

The change in feedwater temperature from

70°F to 40°F maintains the margin of safety

because the replacement steam generators

have been shown by the design analysis

report to be able to withstand the same

number of cycles of the addition of 40°F

water as the old steam generators could 70°F

water.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: Cynthia A.

Carpenter, Acting

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan

Date of amendment request:

September 13, 1993

Description of amendment request:

The proposed amendment would

relocate audit frequencies of Section

6.5.2.8 of the Technical Specifications

to the Quality Assurance Program in

Section 17.2 of the Updated Final Safety

Analysis Report.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change to relocate the audit

program frequency requirements to the

Quality Assurance Program does not:

(1) involve a significant increase in the

probability or consequences of an accident

previously evaluated,

This change is administrative in nature and

does not impact the operation of the plant or

the plant’s response to an accident. Because

it will allow more flexibility in assigning

resources to assess weak or declining

performance areas, the plant safety

performance will be improved.

(2) create the possibility of a new or

different kind of accident from any accident

previously evaluated,

This change is administrative in nature and

does not affect the operation or design of the

plant; therefore, there is no change in the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) involve a significant reduction in a

margin of safety.

This change is administrative in nature and

does not affect the operation of the plant;

therefore, there is no change in the margin of

safety. Relocating the audit program

frequency requirements to the Quality

Assurance program will allow a more

dynamic and responsive audit program.

Audits will be able to be scheduled more

effectively based on performance and the

status of related activities. This should result

in a more effective audit program that will

contribute to an improvement in safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226

NRC Project Director: Cynthia A.

Carpenter, Acting

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, ArkansasNuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request: August

30, 1994, with supplement dated

January 19, 1995.

Description of amendment request:

The proposed amendment changes

requirements related to the site

parimeter security system.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, excerpts of this analysis

are presented below:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated

The accident mitigation features of the

plant are not affected by the proposed

change. This change provides an equivalent

level of protection as required by

10CFR73.55(c)(4), does not significantly

decrease the effectiveness of the security

program, and is adequate for preventing an

unacceptable risk to public health and safety.

Ample protection against a design basis

security threat continues to be provided.

Therefore, the probability or consequences of

an accident previously evaluated is not

significantly increased.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

Any Previously Evaluated

This change clarifies the existing

configuration of the protected area barrier at

the ANO intake structure. New systems,

modes of equipment operation, failure

modes, or other plant perturbations are not

introduced by this change. Therefore, the

possibility of a new or different kind of

accident from amy previously evaluated is

not created.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety

This change clarifies the existing

configuration of the protected area barrier at

the ANO intake structure. The proposed

change does not alter a safety limit, a limiting

condition of operation, or a surveillance

requirement on equipment to operate the

plant. Adequate physical protection of the

plant is maintained. Therefore, the margin of

safety is not significantly reduced.
18626 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

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

Duane Arnold Energy Center (DAEC),

Linn County, Iowa

Date of amendment request: March 1,

1995

Description of amendment request:

The proposed License Amendment

would revise Technical Specification

(TS) Sections 4.5 and 4.8 of the DAEC

TS to reflect the changes to pump and

valve testing criteria. The proposed

amendment changes the testing

frequency for certain pumps and valves

in the Low Pressure Coolant Injection

subsystem; Core Spray subsystems; and

the Residual Heat Removal Service

Water, High Pressure Coolant Injection,

Emergency Service Water, and River

Water Supply systems. The frequency

would change from testing every three

months to that specified by DAEC

ASME Section XI Inservice Testing (IST)

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 affected pumps and valves in

Sections 4.5 and 4.8 will continue to be

tested in accordance with ASME Section XI

OM-6 and OM-10. The affected pumps and

valves will continue to function as before and

this change will not result in a decrease in

their availability to mitigate the

consequences of certain accidents and

transients. The proposed amendment will not

affect the consequences of these accidents

and transients. Therefore, the

proposed amendment does not involve a

change in the probability or consequences of

an accident previously evaluated.

(2) The proposed license amendment does

not create the possibility of a new or different

kind of accident from any previously

evaluated. The safety functions of the

affected pumps and valves will remain

unchanged. This amendment will result in

no physical changes to the affected pumps,

valves or systems. Consequently, the

proposed license amendment does not create

the possibility of a new or different kind of

accident from any previously evaluated.

(3) The proposed amendment will not

reduce the margin of safety. The actual

operation of the affected pumps and valves

will remain unchanged. Testing in

accordance with ASME Section XI OM-6 and

OM-10 will continue to provide assurance

that degradation in tested components will

be detected and addressed.

The NRC staff has reviewed the

licensee’s analysis and, based on

thisreview, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036-5869NRC Acting

Project Director: Gail H. Marcus

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: January

24, 1995, as supplemented March 22,

and March 29, 1995.

Description of amendment request:

The amendment request would revise

the Technical Specification Section

3.2.3.1.a and Table 2.2-1 to decrease the

acceptance criterion for measured

reactor coolant system (RCS) flow rate

from 387,480 gallons per minute (gpm)

to 371,920 gpm.

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 (SHC), which is presented

below:

...The proposed changes do not involve an

SHC because the changes would not:

1. Involve a Significant Increase in the

Probability or Consequence of an Accident

Previously Evaluated.

An evaluation of the 4% decrease in the

RCS total flow rate limit has shown that the

change does not significantly impact the

design basis analyses. Therefore, the change

will not increase the consequences of an

accident previously evaluated.

There are no actual plant changes that will

result from this technical specification

change. Instead, the technical specification

requirement for minimum total RCS flow rate

is being changed to provide operational

benefit without compromising safety. Since

there are no plant changes, there is no effect

on the probability of occurrence of

previously evaluated accidents.

The change will have a negligible impact

on the small break loss of coolant accident

(LOCA) and large break LOCA analyses. The

PCT [peak cladding temperature] acceptance

criteria will continue to be met with the

assumption of a 4% reduction in RCS flow

rate.

For the steam generator tube rupture event,

both the FSAR [Final Safety Analysis Report]

offsite dose analysis and the margin of steam

generator (SG) overfill were evaluated. It was

determined that the 4% reduction in RCS

flow rate will not adversely affect the offsite

doses or the margin to SG overfill and,

therefore, the FSAR conclusions remain

unchanged.

In the evaluation of non-LOCA transients,

the DNB [departure from nucleate boiling] is

the most affected parameter due to a change

in flow rate. It was concluded that the 4%

reduction in RCS flow was acceptable and

there was margin to the DNB limit.

It is concluded that there is sufficient

margin to the system pressure, PCT and DNB

limits to offset the effect of the 4% flow rate

decrease and the calculated radiological

releases associated with the analysis are not

affected. Therefore, there is no effect on the

consequences of previously evaluated

accidents.

2. Create the Possibility of a New or

Different Kind of Accident from any

Previously Evaluated.

The low loop flow trip setpoint specified

in Technical Specification Table 2.2-1 is set

as a fraction of total flow. The flow fraction

is not being changed and no hardware

changes are required due to the reduction in

minimum flow. Also, the reduction in

minimum flow will not change the operation

of any plant equipment and it does not

modify plant operation.

Therefore, the reduction in minimum flow

does not introduce any new failure modes or

malfunctions and it does not create the

potential for a new unanalyzed accident.

3. Involve a Significant Reduction in the

Margin of Safety.

The proposed 4% decrease in the technical

specification limit for total RCS flow rate will

not adversely affect the results of the FSAR

accident analysis, and it is concluded that

this change is safe. The change does not

adversely affect any equipment credited in

the safety analysis, and it does not affect the

probability of occurrence of any plant

accident. Also, the change has a negligible

impact on the PCT, and it does not increase

the offsite doses or decrease the DNB below

its acceptance limit.

Therefore, the change does not have any

significant impact on the protective

boundaries, and there is no reduction in the

margin of safety as specified in the technical

specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18627

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: March 1,

1995

Description of amendment request:

The proposed amendment to the

technical specifications (TS) would

make administrative changes to TS 2.5,

2.8, 2.11, 3.2, and 3.10 and, in

accordance with Generic Letter (GL) 93-

07, ‘‘Modification of the Technical

Specification Administrative Control

Requirements for Emergency and

Security Plans,’’ to TS 5.5 and 5.8.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed revisions to Technical

Specifications (TS) 5.5 and 5.8 are

administrative in nature and follow the

guidance of Generic Letter (GL) 93-07. The

review and audit functions of the site

security and emergency plans and

procedures will be retained in a manner that

fully satisfies regulatory requirements.

Therefore, the proposed revisions do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revision to TS 2.5 will still

require backup water for the emergency

feedwater storage tank to be available.

However, several other available sources of

water are preferred over river water, such as,

the water plant demineralized water system

and the outside condensate storage tank.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed deletion of TS 2.8(8)

pertaining to fuel handling cranes, deletion

of TS 2.11 pertaining to overhead cranes in

the Containment and Auxiliary Buildings,

and deletion of statements in the bases of TS

2.8 pertaining to crane interlocks does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Specifications 2.8(8),

2.11 and the deleted statements in the bases

of Specification 2.8 need not be retained in

the TS based upon Criteria 1 through 4 of the

‘‘Final Policy Statement on Technical

Specifications Improvements for Nuclear

Power Reactors,’’ dated July 22, 1993 (58 FR

39132).

Controls and limitations for the operation

and testing of these cranes and interlocks

will be incorporated into the Updated Safety

Analysis Report (USAR). The requirements of

TS 2.8(8) and restrictions of TS 2.11 are

currently contained in Station procedures to

ensure that the handling of fuel assemblies,

control element assemblies (CEAs) and heavy

loads is accomplished safely and effectively.

These revisions make the FCS Technical

Specifications more similar to Standard

Technical Specifications (STS), which do not

contain requirements or restrictions

concerning the operation of fuel handling

cranes or overhead cranes.

The revision proposed for TS 3.2, Table 3-

5, Item 1 will make its surveillance frequency

identical to the frequency specified in STS

3.1.5.7. The proposed frequency will require

testing CEA drop times prior to reactor

criticality after each removal of the reactor

vessel closure head, which is the most

appropriate time to perform the surveillance.

The proposed frequency will ensure that the

CEAs drop into the core within the time

specified in the safety analysis and, therefore,

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed deletion of TS 3.2, Table 3-

5, Item 5, which currently requires testing

refueling system interlocks prior to the

refueling outage does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. Table 3-5, Item 5, does not need

to be retained in the TS based upon Criteria

1 through 4 of the ‘‘Final Policy Statement on

Technical Specifications Improvements for

Nuclear Power Reactors,’’ dated July 22,

1993. Controls and limitations for testing the

refueling system interlocks will be

incorporated into the USAR. The

requirements for testing refueling system

interlocks are already contained in Station

procedures. This revision makes the FCS

Technical Specifications more similar to

STS, which do not contain requirements or

restrictions pertaining to testing refueling

system interlocks.

The proposed revision to TS 3.2, Table 3-

5, Item 10, ensures consistent use of

terminology among the frequencies specified

in Table 3-5. The proposed revision clarifies

the wording and introduces additional

operational flexibility such that the

surveillance could be performed before 720

hours of system operation, if warranted by

plant conditions or beneficial to plant

operation. Therefore, the proposed revision

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The remaining TS revisions are

administrative in nature in that they correct

references, titles, misspelling(s), and page

numbers, or revise wording to be consistent

with defined intervals within the TS.

Therefore, they do not increase the

probability or consequences of an accident

previously evaluated. None of the proposed

TS revisions will impact the function or

method of operation of plant systems,

structures, or components.

(2) The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed revisions to TS 5.5 and 5.8

which delete the review and/or audit of the

emergency, site security and safeguards

contingency plans and implementing

procedures from the TS are administrative in

nature and in accordance with the guidance

of GL 93-07. The proposed revisions will not

affect the operation of any system, structure,

or component and therefore do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed revision to TS 2.5 will still

require a backup supply of water for the

emergency feedwater storage tank to be

available. However, several other available

sources of water are preferred over river

water, such as, the water plant demineralized

water system and the outside condensate

storage tank. Therefore, the proposed

revision does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

The proposed deletion of TS 2.8(8)

pertaining to fuel handling cranes, deletion

of TS 2.11 pertaining to overhead cranes in

the Containment and Auxiliary Buildings

and deletion of statements in the bases of TS

2.8 pertaining to crane interlocks does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. Specifications 2.8(8),

2.11 and the deleted statements in the bases

of Specification 2.8 need not be retained in

the TS based upon Criteria 1 through 4 of the

‘‘Final Policy Statement on Technical

Specifications Improvements for Nuclear

Power Reactors,’’ dated July 22, 1993.

The requirements of TS 2.8(8) and

restrictions of TS 2.11 are currently

contained in Station procedures to ensure

that the handling of fuel assemblies, CEAs

and heavy loads is accomplished safely and

effectively. These revisions make the FCS

Technical Specifications more similar to

STS, which do not contain requirements or

restrictions concerning the operation of fuel

handling cranes or overhead cranes.

The proposed revision to TS 3.2, Table 3-

5, Item 1, is an administrative revision to the

frequency of CEA drop time testing. The

proposed frequency is the most appropriate

time to perform the surveillance to ensure

that the CEAs drop into the core within the

time specified in safety analysis and is

identical to the frequency specified in STS

3.1.5.7. Therefore, the proposed revision does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed deletion of TS 3.2, Table 3-

5, Item 5, which currently requires testing the

refueling system interlocks prior to the

refueling outage, does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. Table 3-5, Item 5, does not need

to be retained in the TS based upon Criteria

1 through 4 of the ‘‘Final Policy Statement on

Technical Specifications Improvements for

Nuclear Power Reactors,’’ dated July 22,

1993. The requirements for testing refueling
18628 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

system interlocks are currently contained in

Station procedures. This revision makes the

FCS Technical Specifications more similar to

STS, which do not contain requirements or

restrictions pertaining to testing refueling

system interlocks.

The proposed revision to TS 3.2, Table 3-

5, Item 10, ensures consistent use of

terminology among the frequencies specified

in Table 3-5. The proposed revision clarifies

the wording and introduces additional

operational flexibility such that the

surveillance could be performed before 720

hours of system operation, if warranted by

plant conditions or beneficial to plant

operation. Therefore, the proposed revision

does not create the possibility of a new or

different kind of accident from any

previously evaluated.

The remaining TS revisions are

administrative in nature in that they correct

references, titles, misspelling(s), and page

numbers, or revise wording to be consistent

with defined intervals within the TS.

Therefore, they do not create the possibility

of a new or different kind of accident.

(3) The proposed changes do not involve

a significant reduction in a margin of safety.

The proposed revisions to TS 5.5 and 5.8

concerning the review and/or audit of the

emergency, site security and safeguards

contingency plans and implementing

procedures do not involve a significant

reduction in a margin of safety. The audit

and review processes are administrative

functions which will be retained outside the

TS in a manner that fully satisfies regulatory

requirements.

Removing the requirement of TS 2.5 that

Missouri River water from the fire water

system shall be available to provide a backup

water supply to the emergency feedwater

storage tank improves operational flexibility

without reducing any safety margins. Better

sources of backup water are available to

replenish the emergency feedwater storage

tank. Although deleted from TS 2.5, the fire

water system is still required to be available

to meet the requirements of paragraph 3.F of

the FCS Operating License. Therefore, the

proposed revision does not involve a

significant reduction in a margin of safety.

The proposed deletion of TS 2.8(8)

pertaining to fuel handling cranes, deletion

of TS 2.11 pertaining to overhead cranes in

the Containment and Auxiliary Buildings

and deletion of statements in the bases of TS

2.8 pertaining to crane interlocks does not

involve a significant reduction in a margin of

safety. Specifications 2.8(8), 2.11 and the

deleted statements in the bases of

Specification 2.8 do not need to be retained

in the TS based upon Criteria 1 through 4 of

the ‘‘Final Policy Statement on Technical

Specifications Improvements for Nuclear

Power Reactors,’’ dated July 22, 1993.

The requirements of Specification 2.8(8)

and restrictions of Specification 2.11 are

currently contained in Station procedures to

ensure that the handling of fuel assemblies,

CEAs and heavy loads is accomplished safely

and effectively. These revisions make the

FCS Technical Specifications more similar to

STS, which do not contain requirements or

restrictions concerning the operation of fuel

handling cranes or overhead cranes.

The proposed revision to TS 3.2, Table 3-

5, Item 1, is an administrative revision to the

frequency of CEA drop time testing. The

proposed frequency is the most appropriate

time to perform the surveillance to ensure

that the CEAs drop into the core within the

time specified in the safety analysis and is

identical to the frequency specified in STS

3.1.5.7. Therefore, the proposed revision does

not involve a significant reduction in a

margin of safety.

The proposed deletion of TS 3.2, Table 3-

5, Item 5, which currently requires testing the

refueling system interlocks prior to the

refueling outage does not involve a

significant reduction in a margin of safety.

Table 3-5, Item 5, does not need to be

retained in the TS based upon Criteria 1

through 4 of the ‘‘Final Policy Statement on

Technical Specifications Improvements for

Nuclear Power Reactors,’’ dated July 22,

1993. The requirements for testing refueling

system interlocks are currently contained in

Station procedures. This revision makes the

FCS Technical Specifications more similar to

STS, which do not contain requirements or

restrictions pertaining to testing refueling

system interlocks.

The proposed revision to TS 3.2, Table 3-

5, Item 10, ensures consistent use of

terminology among the frequencies specified

in Table 3-5. The proposed revision clarifies

the wording and introduces additional

operational flexibility such that the

surveillance could be performed before 720

hours of system operation if warranted by

plant conditions or beneficial to plant

operation. Therefore, the proposed revision

does not involve a significant reduction in a

margin of safety.

The remaining TS revisions are

administrative in nature in that they correct

references, titles, misspelling(s), and page

numbers, or revise wording to be consistent

with defined intervals within the TS.

Therefore, they do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: LeBoeuf, Lamb,

Leiby, and MacRae, 1875 Connecticut

Avenue, NW., Washington, DC 20009-

5728

NRC Project Director: William H.

Bateman

Southern Nuclear Operating Company,

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

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request: March

6, 1995

Description of amendments request:

The proposed amendment would

relocate the seismic and meteorological

monitoring instrumentation from the

Technical Specifications to the Final

Safety Analysis Report in accordance

with the ‘‘Final Policy Statement on

Technical Specifications Improvements

for Nuclear Power Reactors,’’ dated July

22, 1993.

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 proposed change involve

a significant increase in the probability

or consequences of an accident

previously evaluated?

No. The proposed change relocates

information from the TS to the FSAR and has

no impact on physical plant operation or

configuration. The continued capability of

the seismic and meteorological

instrumentation to perform its intended

function will be ensured through controlled

change processes governed by 10 CFR 50.59.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The sole function of the seismic and

meteorological monitoring instrumentation is

to record data. The proposed change will not

involve any design change or modification to

the plant. The proposed change will not alter

the operation of the plant or the manner in

which it is operated. Any subsequent change

to the Seismic and Meteorological

Monitoring Instrumentation requirements

will undergo a review in accordance with the

criteria of 10 CFR 50.59 to endure that the

change does not involve an unreviewed

safety question.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety?

No. The proposed change will relocate

Seismic and Meteorological Monitoring

Instrumentation requirements from the TS to

licensee controlled documents subject to the

criteria of 10 CFR 50.59. The proposed

change will have no adverse impact on any

protective boundary or safety limit.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18629

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: William H.

Bateman

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request:

November 15, 1994; superseded March

7, 1995 (TS 94-12).

Description of amendment request:

The proposed change would remove the

frequency for each of the audits

specified in the administrative controls

section of the technical specifications

(TS), except those related to the fire

protection system. The requirements to

perform the audits would be retained,

but the frequency for their performance

would be controlled by a requirement to

be added to the Nuclear Quality

Assurance Plan. This would require that

the audits listed in the TS (except those

related to the fire protection system) be

performed on a biennial frequency. In

addition, the proposed change would

remove the requirement to perform site

Radiological Emergency Plan, Physical

Security Plan, and the Safeguard

Contingency Plan reviews and audits

from the TS, since these requirements

presently exist in their respective Plans.

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 standards used to arrive at a

determination that a Technical Specification

change request involves no significant

hazards consideration are included in the

Commission’s regulations, 10 CFR 50.92,

which states that no significant hazards

considerations are involved if the operation

of the facility in accordance with the

proposed amendment would not: (1) involve

a significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of a

new or different kind of accident from any

accident previously evaluated; or (3) involve

a significant reduction in a margin of safety.

Each standard is addressed as follows:

1. Operation of the facility in accordance

with the proposed technical specifications

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The likelihood that an accident will occur

is neither increased or decreased by the

Technical Specification change which only

affects review and audit frequencies. This

Technical Specification change will not

impact the function or method of operation

of plant equipment. Thus, there is not a

significant increase in the probability of a

previously analyzed accident due to this

change. No systems, equipment, or

components are affected by the proposed

changes. Thus, the consequences of a

malfunction of equipment important to safety

previously evaluated in the FSAR are not

increased by this change.

The proposed change only affects review

and audit frequencies. As such, the proposed

change has no impact on accident initiators

or plant equipment, and thus, does not affect

the probabilities or consequences of an

accident.

Therefore, we conclude that this change

does not significantly increase the

probabilities or consequences of an accident.

2. Operation of the facility in accordance

with the proposed technical specifications

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not involve

changes to the physical plant or operations.

Since program audits do not contribute to

accident initiation, a change related to audit

functions cannot produce a new accident

scenario or produce a new type of equipment

malfunction. Also, this change does not alter

any existing accident scenarios. The

proposed change does not affect equipment

or its operation, and, thus, does not create the

possibility of a new or different kind of

accident. Therefore, the proposed change

does not create the possibility of a new or

different kind of accident.

3. Operation of the facility in accordance

with the proposed technical specifications

would not involve a significant reduction in

a margin of safety.

The proposed change concerning conduct

of reviews and audits does not directly affect

plant equipment or operation. Safety limits

and limiting safety system settings are not

affected by this proposed change.

Therefore, use of the proposed Technical

Specification would not involve any

reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Virginia Electric and Power Company,

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

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request: March 2,

1995

Description of amendment request:

The proposed changes would revise

Technical Specification 4.6.1.2.a to

reference the testing requirements of 10

CFR Part 50, Appendix J, and to state

that the Nuclear Regulatory

Commission-approved exemptions to

the applicable regulatory requirements

are permitted.

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:

A discussion of these standards as they

relate to this ... amendment request follows.

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The proposed change ... revises the North

Anna Units 1 and 2 Technical Specification

Surveillance Requirement 4.6.1.2.a to

reference the testing frequency requirements

of 10 CFR 50 Appendix J and to state that

NRC approved exemptions to the applicable

regulatory requirements are permitted. The

current Technical Specification requires

Type A tests be conducted in accordance

with Appendix J to 10 CFR 50. The proposed

administrative change simply includes the

statement ‘‘as modified by NRC-approved

exemptions.’’ No new requirements are

added, nor are any existing requirements

deleted. Any specific changes to the

requirements of Appendix J will require a

submittal from Virginia Electric and Power

Company under 10 CFR 50.12 and

subsequent review and approval by the NRC

prior to implementation. The proposed

change is stated generically to avoid the need

for further Technical Specification changes if

different exemptions are approved in the

future.

The proposed change, in itself, does not

affect reactor operations or accident analyses

and has no radiological consequences. The

change provides clarification so that future

Technical Specifications changes will not be

necessary to correspond to applicable NRCapproved exemptions from the requirements

of Appendix J. This exemption request is

consistent with the intent of the regulation.

Therefore, this proposed change does not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed Technical Specification

amendment for Units 1 and 2 provides

clarification to a specification that

paraphrases a codified requirement.

Since the ... proposed Technical

Specifications change would not change the
18630 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

design, configuration, or method of operation

of the plant, the changes would not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed North Anna Units 1 and 2

Technical Specifications change is

administrative and clarifies the relationship

between the requirements of Technical

Specification Surveillance Requirement

4.6.1.2.a, Appendix J, and any approved

exemptions to Appendix J. It does not, in

itself, change a Safety Limit or a Limiting

Condition for Operation. The NRC will

directly approve any proposed change or

exemption to Appendix J prior to

implementation.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Virginia Electric and Power Company,

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

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request:

November 10, 1994

Description of amendment request:

The proposed amendment request will

clarify the surveillance requirements for

the reactor protection and the

engineered safeguards system

instrumentation and actuation logic.

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:

Operation of Surry Power Station in

accordance with the proposed Technical

Specifications change will not:

1. Involve a significant increase in the

probability of occurrence or consequences of

an accident previously evaluated.

The proposed change to clarify the

surveillance requirements for the Reactor

Protection and Engineered Safeguards

Systems instrumentation and actuation logic

has no impact on the probability of an

accident occurrence. The instrumentation

and actuation logic will continue to be

operated in the same manner. The actual test

frequency is not changing. Rather,

surveillance requirements are being clarified

to represent the actual testing and the

licensing and design bases. Testing of these

instruments and actuation logic are presently

design limited and would otherwise require

using temporary modifications to complete

the testing. Since the testing is not changing,

the clarification of the actual testing does not

contribute to the probability of any

previously analyzed accident. The Reactor

Protection and Engineered Safeguards

Systems instrumentation and actuation logic

will be operated in the same manner and the

system operability requirements are not being

altered. Therefore, the consequences of any

design basis accident are not being increased

by the proposed change to clarify the

surveillance test requirements for the Reactor

Protection and Engineered Safeguards

System instrumentation and actuation logic.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

There are no plant modifications or

changes in methods of plant operation

introduced by this change in the clarification

of the testing for the Reactor Protection and

Engineered Safeguards Systems

instrumentation and actuation logic. The

plant is not being operated or tested in a

different manner due to the proposed change.

Therefore, no new accidents or accident

precursors are generated by the proposed

change to clarify the surveillance test

requirements.

Clarifying the surveillance test

requirements to represent the original

licensing design basis and test conditions

does not create the possibility of a new or

different accident than previously analyzed.

3.Involve a significant reduction in a

margin of safety.

Clarification of the testing for the Reactor

Protection and Engineered Safeguards

Systems instrumentation and actuation logic

surveillance requirements does not affect the

margin of safety in that the operability

requirements for these safety systems remain

unchanged. The existing testing is performed

in accordance with plant design and

licensing basis and provides adequate

indication of the operability of the affected

instrumentation or actuation logic. The

Reactor Protection and Engineered

Safeguards Systems instrumentation and

actuation logic are fully tested on a refueling

cycle basis which includes complete

operation of each relay and end device.

Therefore, the margin of safety is not altered

by the proposed clarification of the testing for

the Reactor Protection and Engineered

Safeguards Systems instrumentation and

actuation logic.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Virginia Electric and Power Company,

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

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request:

November 22, 1994

Description of amendment request:

The proposed amendment request

would delete unnecessary descriptive

phrases regarding the number of cells in

the station and emergency diesel

generator batteries.

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 deletion of the descriptive references

regarding the number of cells in the station

and emergency diesel generator batteries is

an administrative change and therefore does

not:

1. Involve an increase in the probability of

occurrence or consequences of an accident

previously evaluated.

The proposed change to delete the

descriptive references associated with the

station and emergency diesel generator

batteries (60 cell or 56 cell, respectively) has

no impact on the probability of an accident

occurrence. The change is administrative in

nature and therefore does not affect the

operation of the units. The batteries will

continue to be operated in the same manner

as before the change with operability based

on design voltage and capacity requirements

necessary to ensure safety functions can be

performed. Prescribed surveillance testing

will continue to ensure the operability of

individual battery cells. Consequently, the

proposed change does not contribute to the

probability of occurrence or consequences of

any design basis accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This is an administrative change to delete

the descriptive references associated with the

station and emergency diesel generator

batteries. There are no plant modifications

being implemented by the proposed change

and plant operations are not being changed.

Provided the required design voltage and

capacity are maintained, the batteries remain

fully operable and capable of performing

their intended safety functions. Individual

battery cell surveillance requirements remain

unchanged. Therefore, no new accidents or

accident precursors are created by the

proposed change.

3. Involve a reduction in a margin of safety

as defined in the Technical Specifications.
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18631

The proposed administrative change to

delete the descriptive references associated

with the station and emergency diesel

generator batteries (60 cell or 56 cell,

respectively) is administrative in nature.

Provided the required design voltage and

capacity are maintained, the batteries remain

fully operable and capable of performing

their intended safety functions as assumed in

the safety analyses. Individual battery cell

surveillance requirements remain

unchanged. Therefore, the analyzed margin

of safety is not reduced 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 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: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Virginia Electric and Power Company,

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

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: January

24, 1995

Description of amendment request:

The proposed amendment request

would increase the current Technical

Specification pressurizer safety valve

lift setpoint acceptance criterion from

plus or minus 1% as-found and plus or

minus 1% as-left to plus or minus 3%

as-found and plus or minus 1% as-left.

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 Technical Specifications

change does not involve a significant hazards

consideration because operation of Surry

Units 1 and 2 in accordance with this change

would not:

a. involve a significant increase in the

probability or consequences of an accident

previously evaluated. Affected safety-related

parameters were analyzed for a change to

Surry Units 1 and 2 Technical Specification

3.1.A.3.b. It was determined that the primary

and secondary side overpressure safety limits

would not be exceeded in the most limiting

overpressure transient (Loss of Load, Locker

Rotor, and Rod Withdrawal events) with the

pressurizer safety valve lift setpoint

acceptance criterion increased to [plus or

minus] 3%. The DNBR [departure from

nucleate boiling ratio] results of transients

impacted by the setpoint acceptance criterion

increase are not affected by the proposed

change. The increased setpoint acceptance

criterion will not result in an inadvertent

opening of the pressurizer safety valves.

Since the proposed change involves no

alterations to the physical plant, the

probability of occurrence of an accident or

malfunction of equipment important to safety

previously evaluated is not increased.

b. create the possibility of a new or

different kind of accident from any accident

previously identified. The proposed change

to Surry Units 1 and 2 Technical

Specification 3.1.A.3.b does not involve any

alterations to the physical plant which would

introduce any new or unique operational

modes or accident precursors. Only the

allowable tolerance about the existing

setpoint will be changed.

c. involve a significant reduction in a

margin of safety. It was determined that the

most limiting overpressure transients do not

result in maximum pressures in excess of the

primary and secondary side overpressure

limits. The DNBR results of affected

transients are not made more limiting by the

proposed setpoint tolerance increase.

Therefore, the margin of safety is unchanged

by the proposed increase in the safety valve

setpoint acceptance criterion.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

21, 1995

Description of amendment request:

The amendment would revise

Surveillance Requirement 4.6.2.1.d for

the containment spray system to change

the surveillance interval for the

performance of the air or smoke test

through the containment spray header

from once per 5 years to once per 10

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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed reduced testing frequency of

the Containment Spray System nozzles does

not change the way the system is operated or

the Containment Spray System’s operability

requirements. The proposed change to the

surveillance frequency of safety equipment

has no impact on the probability of an

accident occurrence nor can it create a new

or different type of accident. NUREG-1366

concluded that the corrosion of stainless steel

piping is negligible during the extended

surveillance interval. Since the Containment

Spray System is maintained dry there is no

additional mechanism that could cause

blockage of the spray nozzles. Thus, the

nozzles in the Containment Spray System

will remain operable during the ten year

surveillance interval to mitigate the

consequence of an accident previously

evaluated. No clogging or blockage of the

nozzles in the Containment Spray System has

been discovered during the performance of

the five year surveillance tests. Therefore, the

testing of the Containment Spray System[’]s

nozzles at the proposed reduced frequency

will not increase the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed reduced frequency testing of

the Containment Spray System nozzles does

not change the way the Containment Spray

System is operated. The reduced frequency of

testing of the spray nozzles does not change

plant operation or system readiness. The

reduced frequency testing of the Containment

Spray System nozzles does not generate any

new accident precursors. Therefore, the

possibility of a new or different kind of

accident from any accident previously

evaluated is not created by the proposed

changes in surveillance frequency of the

Containment Spray System nozzles.

3. The proposed change does not involve

a significant reduction in the margin of

safety.

Reduced testing of the Containment Spray

System nozzles does not change the way the

system is operated or the Containment Spray

System’s operability requirements. NUREG1366 concluded that the corrosion of

stainless steel piping is negligible during the

extended surveillance interval. Since the

Containment Spray System is maintained dry

there is no additional mechanism that could

cause blockage of the Containment Spray

System nozzles. Thus, the proposed reduced

testing frequency is adequate to ensure spray

nozzle operability. The surveillance

requirements do not affect the margin of

safety in the operability requirements of the

Containment Spray System remains

unaltered. The existing safety analysis

remains bounding. Therefore no margins of

safety are adversely affected by this 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
18632 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

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: March

24, 1995

Description of amendment request:

The proposed amendment would add a

new action statement to Technical

Specification 3.5.1 which would

provide a 72-hour allowed outage time

(AOT) for one accumulator to be

inoperable because its boron

concentration did not meet the 2300-

2500 parts per million (ppm) band. The

amendment would also change the

current allowed outage time for other

reasons of inoperability from 1 hour to

24 hours.

Changes to the surveillance

requirements are also proposed to

incorporate the guidance of Generic

Letter 93-05, ‘‘Line-Item Technical

Specifications Improvements to Reduce

Surveillance Requirements for Testing

During Operation.’’ These proposed

changes would base the operability of

the accumulator on the contained water

volume and cover pressure and would

not require verification of the boron

concentration after an accumulator

volume increase, provided the source of

the makeup water is the refueling water

storage tank.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed change does not involve

a significant Increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The overall protection system

performance will remain within the bounds

of the accident analysis documented in

Chapter 15 of the Updated Safety Analysis

Report [USAR], WCAP-1096-P, and WCAP11883 since no hardware changes are

proposed.

The safety injection accumulators are

credited in Section 15.6.5 of the Updated

Safety Analysis Report for large and small

break LOCA [loss-of-coolant accident]. There

will be no effect on these analyses, or any

other accident analysis, since the analysis

assumptions are unaffected and remain the

same as discussed in Section 15.6.5. Design

basis accidents are not assumed to occur

during allowed outage times covered by the

Technical Specifications. As such, the ECCS

[emergency core cooling system] Evaluation

Model equipment availability assumptions

made in Section 15.6.5 remain valid.

The safety injection accumulators will

continue to function in a manner consistent

with the above analysis assumptions and the

plant design basis. As such, there will be no

degradation in the performance of nor an

increase in the number of challenges to

equipment assumed to function during an

accident situation.

The proposed technical specifications

changes do not involve any hardware

changes nor do they affect the probability of

any event initiators. There will be no change

to normal plant operating parameters, ESF

[engineered safety features] actuation

setpoints, accident mitigation capabilities,

accident analysis assumptions or inputs.

Therefore, these changes will not increase the

probability of an accident or malfunction.

The corresponding increase in CDF [core

damage frequency] due to the proposed

change to increase the AOT of the

accumulators from one hour to 24 hours is

insignificant. Pursuant to the guidance in

Section 3.5 of NSAC-125, the proposed

increase in AOT does not ‘‘degrade below the

design basis the performance of a safety

system assumed to function in the accident

analysis,’’ nor does it ‘‘increase challenges to

safety systems assumed to function in the

accident analysis such that safety system

performance is degraded below the design

basis without compensating effects.’’

Therefore, it is concluded that these changes

do not increase the probability of occurrence

of a malfunction of equipment important to

safety.

(2) The proposed change does not create

the possibility of a new or different kind of

accident from any previously evaluated.

The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. This change is administrative in

nature and does not involve any change to

the installed plant systems or the overall

operating philosophy of WCGS [Wolf Creek

Generating Station].

No new accident scenarios, transient

precursors, failure mechanisms, or limiting

single failures are introduced as a result of

these proposed changes. There will be no

adverse effect or challenges imposed on any

safety-related system as a result of these

changes. Therefore, the possibility of a new

or different type of accident is not created.

There are no changes which would cause

the malfunction of safety-related equipment,

assumed to be operable in the accident

analyses, as a result of the proposed technical

specification changes. No new mode failure

has been created and no new equipment

performance burdens are imposed. Therefore,

the possibility of a new or different

malfunction of safety-related equipment is

not created.

(3) The proposed change does not involve

a significant reduction in the margin of

safety.

The proposed change does not involve an

significant reduction in a margin of safety.

There will be no change to the Departure

from Nucleate Boiling Ratio (DNBR)

Correlation Limit, the design DNBR limits, or

the safety analysis DNBR limits discussed in

Bases Section 2.1.1.

As discussed previously, the performance

of the accumulators will remain within the

assumptions used in the large and small

break LOCA analyses, as presented in USAR

Section 15.6.5. Also, there will be no effect

on the manner in which safety limits or

limiting safety system settings are

determined nor will there be any effect on

those plant systems necessary to assure the

accomplishment of protection functions.

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

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18633

Baltimore Gas and Electric Company,

Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2,

Calvert County, Maryland

Date of amendment request: February

24, 1995

Brief description of amendments: The

proposed amendment would revise the

Calvert Cliffs, Unit No. 2, Technical

Specifications (TSs). Specifically, TS

4.G.1.2 would reference 10 CFR Part 50,

Appendix J, directly, and any approved

exemptions to the Type A testing

frequency requirements, rather than

paraphrase the regulation. The proposed

wording is consistent with that used in

NUREG-1432, ‘‘Standard Technical

Specifications - Combustion

Engineering Plants,’’ dated September

1992.Date of publication of individual

notice in Federal Register: March 8,

1995 (60 FR 12789)

Expiration date of individual notice:

April 7, 1995

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: February

23, 1995, as supplemented March 21,

1995.

Description of amendment request:

The proposed amendment would revise

Technical Specifications 3.8.2.1 and

3.8.3.1 to allow installation of a

modification to replace the battery,

main and tie breakers in response to an

Electrical Distribution Systems

Functional Inspection, conducted by the

NRC in July 1991. The existing breaker

arrangement could result in a trip of

both the battery and main breakers if a

fault occurs on one of the 125 VDC

panelboards. The licensee committed to

have these breakers replaced in 1995

with a better coordinated design to

eliminate the concern.Date of

publication of individual notice in

Federal Register: March 8, 1995 (60 FR

12791)

Expiration date of individual notice:

April 7, 1995

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

No. 50-498, South Texas Project, Unit 1,

Matagorda County, Texas

Date of amendment request: March 1,

1995

Description of amendment request:

The proposed amendment would

modify the steam generator tube

plugging criteria in Technical

Specification 3/4.4.5, Steam Generators,

and the allowable leakage for Unit 1 in

Technical Specification 3/4.4.6.2,

Operational Leakage, and the associated

Bases.Date of individual notice in the

Federal Register: March 13, 1995 (60 FR

13478)

Expiration date of individual notice:

April 12, 1995

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

No. 50-498, South Texas Project, Unit 1,

Matagorda County, Texas

Date of amendment request: March 1,

1995

Description of amendment request:

The proposed amendment would

change Technical Specification 3/4.4.5,

Steam Generators, and the associated

Bases to allow the use of an alternate

plugging criteria (known in the industry

as F*) on steam generator tubes that are

defective or degraded within certain

areas within the tubesheet. Date of

individual notice in the Federal

Register: March 13, 1995 (60 FR 13481)

Expiration date of individual notice:

April 12, 1995

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of amendment request: March 9,

1994

Description of amendment request:

The proposed amendment would revise

the Nine Mile Point Nuclear Station,

Unit 2, Technical Specifications (TSs).

Specifically, TS 4.6.1.2.a would be

modified to allow the second Primary

Containment Integrated Leakage Rate

Test (Type A) to be performed at the

fifth refueling outage (RF-05) or 72

months after the first Type A test

instead of the fourth refueling outage

(RF-04) as currently scheduled.

Date of publication of individual

notice in Federal Register: March 23,

1995 (60 FR 15310)

Expiration date of individual notice:

April 24, 1995

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Rochester Gas and Electric

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

Ginna Nuclear Power Plant, Rochester,

New York

Date of application for amendment:

March 13, 1995

Brief description of amendment: The

proposed amendment would revise

Ginna Station Technical Specification

(TS) 4.4.2.4.a to replace specific leakage

testing frequencies for containment

isolation valves. This TS change will

support a proposed Exemption to Title

10 of the Code of Federal Regulations

(10 CFR) Part 50, Appendix J, Section

III.D.3, requested under separate cover

to exempt Type C testing of certain

valves during a 1995 refueling outage.

Date of publication of individual

notice in Federal Register: March 22,

1995 (60 FR 15167)

Expiration date of individual notice:

April 21, 1995

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

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
18634 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

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.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power

Station,Plymouth County,

MassachusettsDate of application for

amendment: November 22, 1994

Brief description of amendment: The

amendment revises the allowable leak

rate for the main steam isolation valves

from the current 11.5 standard cubic

feet per hour (scfh) for each valve, to a

maximum combined main steam line

leak rate of 46 scfh.

Date of issuance: March 22, 1995

Effective date: March 22, 1995

Amendment No.: 160

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 18, 1995 (60 FR 3671)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 22, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power

Station,Plymouth County,

Massachusetts

Date of application for amendment:

September 6, 1994, as supplemented

February 15, 1995.

Brief description of amendment: This

amendment revises Technical

Specifications (TSs) 3.7.B.1.a, 3.7.B.1.c,

3.7.B.1.e, 3.7.B.2.a, and 3.7.B.2.c and

adds Sections 3.7.B.1.f and 3.7.B.2.e.

The additional section requires both

trains of standby gas treatment and

control room high efficiency air

filtration system to be operable for the

initiation of fuel movement. In the event

either train becomes inoperable, the

other train must be demonstrated to be

operable within 2 hours and fuel

handling operations may continue for 7

days with one train inoperable.

Additionally, this change allows one

train to be defined as operable without

its associated emergency power supply,

provided one source of normal power

(startup transformer or unit auxiliary

power) is available.

Date of issuance: March 22, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 161

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53837) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 22, 1995. No significant hazards

consideration comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power

Station,Plymouth County,

Massachusetts

Date of application for amendment:

September 6, 1994

Brief description of amendment: This

amendment would reduce the Reactor

Pressure Setpoint at which the

shutdown cooling system automatically

isolates. This setpoint also isolates the

low pressure coolant injection valves

when the shutdown cooling system is in

operation.

Date of issuance: March 27, 1995

Effective date: To be implemented

within 30 days following restart from

refueling outage ι10

Amendment No.: 162

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53837) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 27, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Carolina Power & Light Company, et

al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County,North

Carolina

Date of application for amendments:

October 28, 1994, as supplemented

February 16, 1995.

Brief description of amendments: The

proposed change will revise TS

requirements to increase the

surveillance test intervals and the

allowable out of service times or

instruments of the reactor protection

system, isolation actuation system,

emergency core cooling system

actuation system, control rod

withdrawal block system, control room

emergency ventilation system,

anticipated transient without scram,

recirculation pump trip (RPT), end-ofcycle RPT, and the reactor core isolation

cooling actuation system.

Date of issuance: March 30,

1995Effective date: March 30, 1995

Amendment Nos.: 175 and 206

Facility Operating License Nos. DPR71 and DPR-62. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63114) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 30, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North CarolinaDate

of application for amendment: October

24, 1994, as supplemented December 6,

1994.

Brief description of amendment: The

amendment allows the relocation of TS

3/4.3.4, Turbine Overspeed Protection

and associated Bases to be consistent

with the new Standard Technical

Specifications for Westinghouse plants.

Date of issuance: March 22, 1995

Effective date: March 22, 1995

Amendment No. 55

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60379) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 22, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18635

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, IllinoisDocket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of application for amendments:

December 22, 1992

Brief description of amendments:

These amendments add new

requirements to the Technical

Specifications (TS) to ensure that an

Essential Service Water system (SX)

pump and crossover path are available

from a shutdown unit to serve as backup

to an operating unit. In addition, a new

TS is added to require the unit crosstie

to be open, or capable of being opened,

from the Main Control Room, whenever

either, or both units are in an operating

mode (MODE 1, 2, 3, or 4).

Date of issuance: March 20, 1995

Effective date: March 20, 1995

Amendment Nos.: 71, 71, 62, and 62

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 3, 1993 (58 FR 6994)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 20, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Byron, the Byron Public

Library, 109 N. Franklin, P.O. Box 434,

Byron, Illinois 61010; for Braidwood,

the Wilmington Township Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Consolidated Edison Company of New

York, Docket No. 50-247, Indian

PointNuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

September 19, 1994

Brief description of amendment: The

amendment would revise Technical

Specification Section 4.4.A.3,

Frequency of Containment Integrated

Leakage Rate Test, to reference 10 CFR

Part 50, Appendix J, as modified by

approved exemptions, directly.

Date of issuance: March 17, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment No.: 181

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8744) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 17, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, PennsylvaniaDate of

application for amendments: April 23,

1990, as supplemented January 21,

1992 and March 17, 1995.

Brief description of amendments:

These amendments revise the Appendix

A Technical Specifications (TSs) for

Unit 1 and Unit 2 by (a) deleting TS

Table 3.6-1, ‘‘Containment

Penetrations,’’ (b) rewording TS

Definition 1.8, ‘‘Containment Integrity,’’

and TSs 3.6.1.1, 3.6.1.2, 3.6.3.1, and

3.9.4 relating to containment integrity,

containment leakage, containment

isolation valves, and containment

building penetrations respectively to

account for the deletion of TS Table 3.6-

1, and (c) correcting terminology by

replacing the word ‘‘door’’ with ‘‘hatch’’

in TS 3.9.4.a.

The Unit 1 amendment also modifies

TS Table 3.3-5, ‘‘Engineered Safety

Features Response Times,’’ by changing

the feedwater isolation response time to

reflect total isolation times for the main

feedwater regulating valve and bypass

feedwater regulating valve. Minor

editorial changes were also incorporated

in TS Table 3.3-5.

Date of issuance: March 28, 1995

Effective date: March 28, 1995

Amendment Nos.: 185 and 66

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

the Technical Specifications.

Date of initial notice in Federal

Register: June 27, 1990 (55 FR 26283),

as supplemented April 1, 1992 (57 FR

11107) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 28, 1995. No significant hazards

consideration comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

June 22, 1994

Brief description of amendment: The

amendment changes Technical

Specification (TS) Sections 1.6, 3.2.A,

3.9.f.5 and 4.2.A which specify the

Shutdown Margin (SDM) requirements

that ensure the reactor can be made

subcritical and can be maintained

sufficiently subcritical to preclude

inadvertent criticality in any core

condition. The amendment also

includes a definition of Shutdown

Margin, TS Section 1.45. Administrative

changes to TS Sections 1.7 and

3.2.b.2(b) are also included to simplify

definitions and eliminate unnecessary

notes and references.

Date of Issuance: March 21,

1995Effective date: As of the date of

issuance to be implemented within60

days

Amendment No.: 178

Facility Operating License No. DPR16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37072)

The Commission’s related evaluation of

this amendment is contained in a Safety

Evaluation dated March 21, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

February 14, 1995

Brief description of amendment: The

amendment revises Technical

Specification 3.8.2, ‘‘AC SourcesShutdown;’’ 3.8.5, ‘‘DC SourcesShutdown;’’ and 3.8.8, ‘‘InvertersShutdown.’’ The changes revise the

operability requirements for the

Division 3 diesel generator and the

Division 3 and 4 batteries, battery

chargers and inverters to apply only

when the high pressure core spray

system is required to be operable.

Date of issuance: March 21, 1995

Effective date: March 21, 1995

Amendment No.: 99

Facility Operating License No. NPF62. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 17, 1995 (60 FR

9412) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 21, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727.
18636 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

Niagara Mohawk Power Corporation,

Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of application for amendment:

June 30, 1994, as supplemented March

7, 1995

Brief description of amendment: The

amendment revises Technical

Specification (TS) 3.2.7.1 to add 8 check

valves to Table 3.2.7.1. These valves

were installed to add additional

protection of the low pressure Core

Spray system from the high pressure

Reactor Coolant system. Including the

valves in the TSs will assure that the

proper surveillance testing is done to

maintain a high reliability for the valves

to protect the Core Spray system.

Date of issuance: March 20, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 154

Facility Operating License No. DPR63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39593)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 20, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

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:

April 22, 1994

Brief description of amendment: The

amendment deletes the operability and

surveillance requirements of the

condenser air ejector radiation monitor

from the Millstone Unit 2 Technical

Specification Tables 3.3-12 and 4.3-12.

Date of issuance: March 27, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment No.: 186

Facility Operating License No. DPR65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27058)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 27, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

March 31, 1994 and August 5, 1994

Brief description of amendment: This

amendment revises: Technical

Specification (TS) 3.8.1.1.b.2 which

maintains diesel operability for a 48-

hour period when the fuel storage

system of one or more diesel generators

contains less than a 7-day supply of

fuel: TS 4.8.1.1.2.h.8 by deletion and

replacement with surveillance

requirement 4.8.1.1.2.k.1 which permits

the 24-hour diesel generator endurance

run to be performed in any operational

condition; establish surveillance

requirement 4.8.1.1.2.k.2 which allows

the hot restart test to be conducted not

only after surveillance requirement

4.8.1.1.2.k.1, but also after the diesel

generator has operated between 4300 kw

and 4400 kw for one hour or after any

time the diesel generator operating

temperature has stabilized; revise TS

3.8.1.1 to eliminate the requirements to

start the Emergency Diesel Generator

(EDG) with an inoperable offsite

circuit(s) of AC electrical power; add a

provision that eliminates required

testing of remaining EDGs when one

EDG is inoperable due to an inoperable

support system or an independently

testable component with no potential

for common mode failure for the

remaining EDGs. In addition, if testing

of the EDGs is required, the surveillance

will be performed within 16 hours

instead of 24 hours as currently

specified; delete the requirement to

perform a Loss of Offsite Power (LOOP)

test (Surveillance Requirement

4.8.1.1.2.h.b) following the 24-hour EDG

endurance run test in its place, a hot

restart test (no LOOP load sequencing)

will be established.

Date of issuance: March 30, 1995

Effective date: March 30,1995

Amendment No.: 72

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 8, 1994 (59 FR 29630) and

October 12, 1994 (59 FR 51625) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated March 30, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

South Carolina Electric & Gas

Company, South Carolina Public

ServiceAuthority, Docket No. 50-395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of application for amendment:

October 29, 1993, as supplemented on

March 11, 1994, May 18, 1994,

September 20, 1994, and October 20,

1994.

Brief description of amendment: The

amendment changes Operating License

NPF-12 to delete License Conditions

2.C.13, 2.C.14, and 2.C.32.

Date of issuance: March 29, 1995

Effective date: March 29, 1995

Amendment No.: 123

Facility Operating License No. NPF12. Amendment revises the operating

license.

Date of initial notice in Federal

Register: February 16, 1994 (59 FR

7698) and April 28, 1994 (59 FR 22012),

as corrected June 30, 1994 (59 FR

33795). The May 18, 1994, September

20, 1994, and October 20, 1994,

submittals provided supplemental and

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

March 29, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Southern California Edison Company,

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

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

September 30, 1993, as supplemented

by letters dated November 16, 1993,

January 18, 1995, and February 2, 1995.

Brief description of amendments:

These amendments revised the

technical specifications to (1) divide

item 7 of Tables 3.3-3, 3.3-4, 3.3-5, and

4.3-2 into item 7a that addresses the

existing loss-of-voltage (LOV) function

and item 7b that separately addresses

the degraded grid voltage (DGV)

function; (2) add footnote (d) to Table

3.3-3 to indicate that the DGV actuation

relay logic is applicable in Modes 1, 2,

3, and 4 when the diesel generator

circuit breaker is open; (3) replace the

reference to Figure 3.3-1 in item 7a of

Tables 3.3-4 and 3.3-5 with definite

voltage and time values; (4) add note 9

to Table 3.3-5 to explain the response
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18637

time for an LOV signal; and (5) delete

Figure 3.3-1, ‘‘Degraded Bus Voltage

Trip Setting,’’ and the reference to this

figure from Table 3.3-4.

Date of issuance: March 17, 1995

Effective date: Unit 2, as of the date

of completion of the currrent refueling

outage and must be fully implemented

before the plant returns to power; Unit

3, as of the date of the completion of its

next refueling outage and must be fully

implemented before the plant returns to

power.

Amendment Nos.: Unit 2 -

Amendment No. 118; Unit 3 -

Amendment No. 107

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 10, 1993 (58 FR

59755). The additional information

contained in the November 16, 1993,

January 18, 1995 and February 2, 1995,

letters was clarifying in nature, within

the scope of the initial notice and did

not affect the NRC staff’s proposed no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

March 17, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713

Southern Nuclear Operating Company,

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

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama.

Date of amendments request: January

9, 1995

Brief description of amendments: The

amendments change the Technical

Specifications to implement

recommended changes from Generic

Letter (GL) 93-05, ‘‘Line Item Technical

Specification Improvements to Reduce

Surveillance Requirements for Testing

During Power Operation,’’ dated

September 27, 1993. Specifically, the

amendments implement TS changes

corresponding to the following GL 93-05

line-item improvement issues and

numbers: Control Rod Movement Test

for Pressurized Water Reactors (4.2.1);

Radiation Monitors (5.14); Surveillance

of Boron Concentration in the

Accumulator/Safety Injection/Core

Flood Tank (7.1); Containment Spray

System (8.1); Hydrogen Recombiner

(8.5); and Special Test Exemptions (12).

Date of issuance: March 20, 1995

Effective date: March 20, 1995

Amendment Nos.: 113 and 104

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

Technical Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8756) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 20, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

December 6, 1994

Brief description of amendment: This

amendment deletes Technical

Specification (TS) Surveillance

Requirement (SR) 4.1.3.2.2 for the Axial

Power Shaping Rods and relaxes

surveillance intervals for TS 3/4.1.3.1,

‘‘Group Height - Safety and Regulating

Rod Groups;’’ TS 3/4.4.6.2, ‘‘Operational

Leakage;’’ TS 3/4.5.2, ‘‘ECCS

Subsystems - Tavg equal to or greater

than 280°F;’’ TS 3/4.6.2.1,

‘‘Containment Spray System;’’ and TS 3/

4.10.4, ‘‘Special Test Exceptions

Shutdown Margin.’’ Date of issuance:

March 21, 1995Effective date: March 21,

1995 and implemented not later than 90

days after issuance

Amendment No.: 196

Facility Operating License No. NPF-3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8757) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 21, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

December 6, 1994

Brief description of amendment: This

amendment revises Technical

Specification (TS) 4.0.5, ‘‘Applicability’’

and its associated Bases to eliminate the

need for NRC approval of relief requests

prior to implementation and relaxes

surveillance test intervals for TS 3/

4.1.2.3, ‘‘Reactivity Control Systems -

Makeup Pump - Shutdown; TS 3/

4.1.2.4, ‘‘Reactivity Control Systems -

Makeup Pumps - Operating; TS 3/

4.1.2.6, Reactivity Control Systems -

Boric Acid Pump - Shutdown; and TS

3/4.1.2.7, ‘‘Reactivity Control System -

Boric Acid Pumps - Operating’’ from

monthly to quarterly. Date of issuance:

March 22, 1995

Effective date: March 22, 1995, and to

be implemented within 90 days

Amendment No.: 197

Facility Operating License No. NPF-3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8758) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 22, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

December 9, 1994, as supplemented on

December 22, 1994.

Brief description of amendment: The

amendment revises the Technical

Specification (TS) Surveillance

Requirement 4.8.1.1.2f.7. The change

removes the requirement to perform the

hot restart test within 5 minutes of

completing the 24-hour endurance test

and places that requirement in a

separate TS.

Date of issuance: March 20, 1995

Effective date: March 20, 1995, to be

implemented within 30 days

Amendment No.: 95

Facility Operating License No. NPF30. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6315)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 20, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.
18638 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

August 4, 1994, as supplemented on

March 14, 1995 and March 28, 1995.

Brief description of amendment: The

amendment replaces Technical

Specification (TS) 3/4.6.2.2, Spray

Additive System, with a new TS 3/

4.6.2.2 entitled Recirculation Fluid pH

control (RFPC) System. The associated

TS Surveillance Requirements and the

Bases will also be revised. In addition,

the Bases section for the Refueling

Water Storage Tank (RWST) System will

be revised.

Date of issuance: March 30, 1995

Effective date: March 30, 1995, to be

implemented within 30 days

Amendment No.: 96

Facility Operating License No. NPF30. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49440) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 30, 1995. The March 14, 1995,

and March 28, 1995, letters provided

supplemental information that did not

change the initial proposed no

significant hazards consideration

determination. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

September 8, 1994

Brief description of amendment: The

amendment revises the Technical

Specification (TS) Bases Section 3/4.9

and changes Final Safety Analysis

Report (FSAR) Sections 9.1.3 ‘‘Fuel Pool

Cooling and Cleanup,’’ 9.1.4 ‘‘Fuel

Handling System’’ and 15.4.6 ‘‘Chemical

and Volume Control System

Malfunction That Results in a Decrease

in the Boron Concentration in the

Reactor Coolant. The changes

established procedural controls to

address an unreviewed safety question.

Date of issuance: March 31, 1995

Effective date: March 31, 1995, to be

implemented within 30 days

Amendment No.: 97

Facility Operating License No. NPF30. Amendment revises the Technical

Specification Bases and FSAR.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11151)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 31, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of application for amendment:

December 8, 1994, as supplemented by

letter dated February 16, 1995.

Brief description of amendment: The

proposed amendment would change

Standby Gas Treatment Power Supply

Requirements during refueling

operations.

Date of issuance: March 23, 1995

Effective date: As of the date of

issuance, to be implemented within 30

days

Amendment No.: 143

Facility Operating License No. DPR28. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8759) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 23, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont

05301.

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

October 31, 1994

Brief description of amendment: The

amendment relocated requirements

regarding safety/relief valve position

indication instrumentation from the

Technical Specifications to other

licensee-controlled documents.

Date of issuance: March 27, 1995

Effective date: March 27, 1995, to be

implemented prior to restart from the

spring 1995 refueling outage

Amendment No.: 135

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65831) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 27, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee

NuclearPower Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

December 2, 1994

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant (KNPP) Technical

Specification (TS) 3.2 by deleting the

requirements for the charging pumps,

high concentration boric acid in the

boric acid storage tanks (BASTs), the

boric acid transfer pumps, and boric

acid heat tracing. Changes to TS 3.3 and

Table TS 3.5.3 add requirements

associated with the emergency core

cooling system (ECCS) accumulators,

remove the requirements associated

with the boric acid storage tanks and

increase the minimum required boron

concentration in the refueling water

storage tank (RWST). Additionally, the

surveillance requirements involving the

BASTs, associated valves and heat

tracing located in Table TS 4.1-1, Table

TS 4.1-2 and Section 4.5 have been

deleted.

Date of issuance: March 28, 1995

Effective date: March 28, 1995, to be

implemented within 20 days

Amendment No.: 116

Facility Operating License No. DPR43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 508).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 28, 1995. No

significant hazards consideration

comments received: None.

Local Public Document Room

location: University of Wisconsin

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301.

Notice Of Issuance Of Amendments To

Facility Operating LicensesAnd Final

Determination Of No Significant

Hazards ConsiderationAnd

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules
Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices 18639

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By May

12, 1995, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to
18640 Federal Register / Vol. 60, No. 70 / Wednesday, April 12, 1995 / Notices

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

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

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

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

Northeast Nuclear Energy Company,

Docket No. 50-245, MillstoneNuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

March 17, 1995

Brief description of amendment: The

amendment revises Technical

Specification (TS) Surveillance

Requirement 4.7.D.1.c.1 by replacing the

once per quarter stroke test for

containment isolation valves (CIVs)

with the requirement that the CIVs be

tested in accordance with the inservice

testing program. In addition, there are

some editorial changes, minor

renumbering of subsections, to reflect

the TS revisions.

Date of issuance: March 21, 1995

Effective date: As of the date of

issuance to be implemented

immediately

Amendment No.: 81

Facility Operating License No. DPR21. Amendment revised the Technical

Specifications. Public comments

requested as to proposed no significant

hazards consideration: No. The

Commission’s related evaluation of the

amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated March 21, 1995.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Dated at Rockville, Maryland, this 5th day

of April, 1995.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 95–8845 Filed 4–11–95; 8:45 am]

BILLING CODE 7590–01–F

OFFICE OF PERSONNEL

MANAGEMENT

Notice of Request for Extension of INV

Forms 40–44 Submitted to OMB for

Clearance

AGENCY: Office of Personnel

Management.

ACTION: Notice.

SUMMARY: In accordance with the

Paperwork Reduction Act of 1980 (title

44, U.S. Code, chapter 35), this notice

announces the reclearance of forms used

to request information by mail for use in

OPM investigations. These

investigations are conducted to

determine suitability for Federal

employment and/or the ability to hold

a security clearance, as prescribed in

Executive Orders 10450 and 10577 and

5 U.S.C. 3301.

INV Form 41, Investigative Request

for Employment Data and Supervisor

Information, is sent to former employers

and/or supervisors; INV Form 42,

Investigative Request for Personal

Information, is sent to references; INV

Form 43, Investigative Request for

Educational Registrar and Dean of

Students Record Data, is sent to

educational institutions; and INV Form

44, Investigative Request for Law

Enforcement Data, is sent to local law

enforcement agencies. In order to

accommodate sources for which the

collection formats of INV Forms 41–44

are awkward or inappropriate, INV

Form 40, General Request for

Investigative Data, has been added to

the collection.

It is estimated that 1,065,955

individuals will respond annually

(186,408 to INV Form 40; 360,115 to

INV Form 41; 284,160 to INV Form 42;

76,152 to INV Form 43; and 159,120 to

INV Form 44), and that each will require

approximately 5 minutes to complete

the form, for a total burden of 88,830

hours. For copies of this proposal call

Doris R. Benz on 703–908–8564.

DATES: Comments on this proposal

should be received on or before May 12,

1995.

ADDRESSES: Send or deliver comments

to—Joseph Lackey, OPM Desk Officer,

Office of Information and Regulatory

Affairs, Office of Management and

Budget, New Executive Office Building,

NW, Room 10235, Washington, DC

20503.

FOR FURTHER INFORMATION CONTACT:

John J. Lafferty, 202–376–3800.

Office of Personnel Management.

James B. King,

Director.

[FR Doc. 95–8975 Filed 4–11–95; 8:45 am]

BILLING CODE 6325–01–M

SECURITIES AND EXCHANGE

COMMISSION

[Rel. Nos. 33–7156; 34–35572]

Changes and Corrections to EDGAR

Phase-In List

AGENCY: Securities and Exchange

Commission.

ACTION: Notice.

SUMMARY: The Commission is

publishing a list of changes and

corrections to the EDGAR phase-in list

for companies whose filings are

processed by the Division of

Corporation Finance.

FOR FURTHER INFORMATION CONTACT:

Sylvia J. Reis, Assistant Director, CF

EDGAR Policy, Division of Corporation

Finance at (202) 942–2940.

SUPPLEMENTARY INFORMATION: In

connection with the adoption of the

final rules fully implementing the