Biweekly Notice

NUCLEAR REGULATORY COMMISSION

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 [[Page 18622]] 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 [[Page 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. [[Page 18624]]

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 NRC-approved

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 deg.F to 40 deg.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 deg.F

to 40 deg.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 deg.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 deg.F water as the old steam generators could with

70 deg.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 deg.F as the auxiliary feedwater temperature

and thus bounds [the] 40 deg.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 deg.F to

40 deg.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 [[Page 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 deg.F to 40 deg.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 deg.F

water as the old steam generators could 70 deg.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. [[Page 18626]]

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

[[Page 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

[[Page 18628]] 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

[[Page 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 NRC-approved 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 [[Page 18630]] 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. [[Page 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. 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 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

[[Page 18632]] 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 WCAP-11883

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. [[Page 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 [[Page 18634]] 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. DPR-35: 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. DPR-35: 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 10

Amendment No.: 162

Facility Operating License No. DPR-35: 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-of-cycle 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. DPR-71 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. NPF-63. 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. [[Page 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. NPF-37, 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. DPR-26: 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. DPR-66 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. DPR-16. 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 Sources-Shutdown;'' 3.8.5, ``DC Sources-

Shutdown;'' and 3.8.8, ``Inverters-Shutdown.'' 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. NPF-62. 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. [[Page 18636]]

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. DPR-63: 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. DPR-65. 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. NPF-57: 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. NPF-12. 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 [[Page 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. NPF-10 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. NPF-2 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 deg.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. NPF-30. 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. [[Page 18638]]

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. NPF-30. 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. NPF-30. 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. DPR-28. 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. NPF-21: 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. DPR-43. 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

[[Page 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 [[Page 18640]] 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. DPR-21. 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