Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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 May 1, 1995, through May 12, 1995. The last

biweekly notice was published on May 10, 1995 (60 FR 24904).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By June 23, 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 [[Page 27335]] bases of the contention and a

concise statement of the alleged facts or expert opinion which support

the contention and on which the petitioner intends to rely in proving

the contention at the hearing. The petitioner must also provide

references to those specific sources and documents of which the

petitioner is aware and on which the petitioner intends to rely to

establish those facts or expert opinion. Petitioner must provide

sufficient information to show that a genuine dispute exists with the

applicant on a material issue of law or fact. Contentions shall be

limited to matters within the scope of the amendment under

consideration. The contention must be one which, if proven, would

entitle the petitioner to relief. A petitioner who fails to file such a

supplement which satisfies these requirements with respect to at least

one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

date. Where petitions are filed during the last 10 days of the notice

period, it is requested that the petitioner promptly so inform the

Commission by a toll-free telephone call to Western Union at 1-(800)

248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator

should be given Datagram Identification Number N1023 and the following

message addressed to (Project Director): petitioner's name and

telephone number, date petition was mailed, plant name, and publication

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,

and 3, Maricopa County, Arizona

Date of application for amendments: April 6, 1995.

Brief description of amendments: The proposed amendment involves

changes in personnel titles, implementation of line item improvements

delineated in Generic Letter 93-07, ``Modification of the Technical

Specification Administrative Control Requirements for Emergency and

Security Plans,'' changes in the Plant Review Board, and miscellaneous

minor changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

These changes involve (1) minor changes in the organization of

PVNGS, (2) line item improvements recommended by the NRC, or (3)

clarification or corrections to existing specifications. It is

expected that the organizational changes will have a positive effect

on the conduct of plant operations and safety-related work.

Functions which are necessary to operate the facility safely and in

accordance with the operating licenses, remain in the new

organization. The line item improvements to the Technical

Specifications will not affect the safe operation of the plant and

continue to ensure proper control of administrative activities. The

proposed changes will not affect the operation of structures,

systems and components, and will not reduce programmatic controls

such that plant safety would be affected. Therefore, the proposed

changes do 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 kind of accident from any accident previously analyzed.

The proposed changes will not affect the operation of

structures, systems and components, and will not reduce programmatic

controls such that plant safety would be affected. The changes in

the organization and as a result of line item improvements will

continue to provide necessary oversight and control of

administrative processes. Therefore, the proposed changes do not

create the possibility of a new or different kind of accident from

any previously evaluated.

(3) The proposed changes do not involve a significant reduction

in a margin of safety.

These changes are administrative and will not diminish any

organizational or administrative controls currently in place. The

proposed changes will not affect the operation of structures,

systems and components, and will not reduce programmatic controls

such that plant safety would be affected. Therefore, the proposed

changes do not involve a significant reduction in a margin of

safety.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.

1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: April 18, 1995.

Description of amendment requests: The proposed Technical

Specification amendments would revise the surveillance requirements for

Technical Specification 3/4.4.4, ``Steam Generators,'' and the

associated Bases. These amendments would allow the installation of tube

sleeves as an alternative to plugging defective steam generator tubes.

Basis for proposed no significant hazards consideration

determination: [[Page 27336]] 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 amendment to permit the use of steam generator tube

sleeves as an alternative to tube plugging is a safe and effective

repair procedure that does not require removing a tube from service.

Mechanical strength, corrosion resistance, installation methods, and

inservice inspection techniques of sleeves have been shown to meet

NRC acceptance criteria.

Analytical verifications were performed using design and

operating transient parameters selected to envelope loads imposed

during normal operating and accident conditions. Fatigue and stress

analysis of sleeved tube assemblies were completed in accordance

with the requirements of Section III of the ASME Code. The results

of qualification testing, analysis and plant operating experience at

other facilities demonstrates that the sleeving process is an

acceptable means of maintaining steam generator tube integrity. The

sleeve configuration has been designed and analyzed in accordance

with the structural margins specified in Regulatory Guide (RG)

1.121. Furthermore, the installed sleeve will be monitored through

periodic inspections on a sample basis with eddy current techniques.

A sleeve-specific plugging margin, per the recommendations of RG

1.121, has been specified with appropriate allowances for NDE

(nondestructive examination) uncertainty and defect growth rate.

The consequences of accidents previously analyzed are not

increased as a result of sleeving activities. The hypothetical

failure of the sleeve would be bounded by the current steam

generator tube rupture analysis contained in the PVNGS (Palo Verde

Nuclear Generating Station) UFSAR (updated final safety analysis

report). Due to the slight reduction in diameter caused by the

sleeve wall thickness, it is expected that the primary release rates

would be less than assumed for the steam generator tube rupture

analysis, and therefore would result in lower total primary fluid

mass release to the secondary system. Additionally, further

conservatism is introduced if the break were postulated to occur at

a location on the tube higher than the location where a sleeve is

installed. The overall effect would be reduced steam generator tube

rupture release rates. The minimal reduction in flow area associated

with a tube sleeve has no significant affect on steam generator

performance with respect to heat transfer or system flow resistance

and pressure drop. The installation of sleeves rather than plugging

also maintains a greater heat transfer surface in the steam

generator. In any case, the impacts are bounded by evaluations which

demonstrate the acceptability of tube plugging which totally removes

the tube from service. Therefore, in comparison to plugging, tube

sleeving is considered a significant improvement with respect to

steam generator performance. The cumulative impact of multiple

sleeved tubes was evaluated to ensure the effects remain within the

analytical design bases.

Recent industry experience with forced shutdown events

associated with tube failures at sleeve junctions was assessed by

ASP and ABB-CE. The root cause of these events has been attributed

to the lack of proper post-installation stress relief and/or the

imposition of high stresses due the tube growth restrictions at

locked tube support. The material and design of the PVNGS steam

generator supports minimizes the potential for locked supports. The

tube supports are of eggcrate design and are constructed of ferritic

stainless steel. The large flow area in the eggcrate design provides

better irrigation and reduces the potential for steam blanketing,

therefore, the tube-to-tube support crevices are less likely to be

blocked by crud, boiler water deposits and corrosion products. Since

the support material is type 409 ferritic stainless steel, it is not

susceptible to magnetite corrosion which has resulted in denting and

lockup at plants with carbon steel supports. These conclusions have

been substantiated via tube pull activities conducted in PVNGS Unit

2. Although ABB-CE does not require post-weld heat treatment in all

applications, APS will require that a post-weld stress relief be

conducted for all sleeve installations.

APS has incorporated an integrated leakage monitoring program,

utilizing equipment, procedure upgrades and administrative shutdown

limits significantly lower than Technical Specification

requirements. The program is designed to provide plant operators

with the ability to detect and respond to changes in primary-to-

secondary leakage and shutdown the unit prior to a significant leak

or steam generator tube rupture, should sleeve or tube degradation

exceed expected values. The program is designed to reduce the

probability of steam generator tube rupture events.

Therefore, based on the above, the proposed amendment does not

significantly increase the probability or consequences of an

accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously analyzed.

A sleeved steam generator tube performs the same function in the

same passive manner as an unsleeved steam generator tube. Tube

sleeves are designed, qualified, and maintained under the stress and

pressure limits of Section III of the ASME Code and Regulatory Guide

1.121.

The installation of the sleeve, including weld and welder

qualification and nondestructive examination (NDE), meets or exceeds

the requirements of ASME Section XI. Three types of NDE are

conducted. Ultrasonic Testing (UT) is performed to verify adequacy

of the tube to sleeve weld assuring proper fusion. Eddy current

testing (ET) is performed following each installation to establish

baseline data for each sleeve in order to monitor future degradation

of the primary to secondary pressure boundary. Visual inspections

may be performed to verify or ascertain the mechanical and

structural condition of a weld. Critical conditions which are

checked include weld width and completeness, and the absence of

visibly noticeable indications such as cracks, pits, and burn

through.

ABB-Combustion Engineering Inc., Report CEN-613-P, ``Arizona

Public Service Co., Palo Verde Units 1, 2, and 3, Steam Generator

Tube Repair Using Leak Tight Sleeves,'' Revision 01, January 1995,

demonstrates that the repair of degraded steam generator tubes using

tube sleeves will result in tube bundle integrity consistent with

the original design basis. An extensive analysis and corrosion and

mechanical test programs were undertaken to prove the adequacy of

tube sleeve repair. The proposed amendments have no significant

effect on the configuration of the plant, and the change does not

effect the way in which the plant is operated. Based upon the

results of the analytical and test programs described in the ABB

Combustion Engineering Inc. report, the tube sleeve fulfills its

intended function and meets or exceeds established design criteria.

Therefore, the proposed change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Evaluation of the sleeved tubes indicates no detrimental effects

on the sleeve-tube assembly resulting from reactor system flow,

coolant chemistries, or thermal and pressure conditions. Structural

analyses of the sleeve-tube assembly, using demonstrated margins of

safety, have established sleeve-tube integrity under normal and

accident conditions. Structural analyses have been performed for

sleeves which span the tube at the top of the tubesheet and which

span the flow distribution plate or eggcrate support. Mechanical

testing has been performed to support the analyses. Corrosion

testing of typical sleeve-tube assemblies has been completed and

reveals no evidence of sleeve or tube corrosion considered

detrimental under anticipated service conditions.

Based upon the testing and analyses performed, the installation

of tube sleeves will not result in a significant reduction in a

margin of safety.

Steam generator tube integrity is maintained under the same

limits for sleeved tubes as for unsleeved tubes, i.e., Section III

of the ASME Code and Regulatory Guide 1.121. The portions of the

installed sleeve assembly which represents the reactor coolant

pressure boundary can be monitored for the initiation and

progression of sleeve/tube wall degradation, thus satisfying the

requirements of Regulatory Guide 1.83. The degradation limit at

which a sleeve/tube boundary is considered inoperable has been

analyzed in accordance with Regulatory Guide 1.121 and is specified.

Eddy current detectability of flaws has been verified by ABB

Combustion Engineering. The Technical Specifications continue to

require monitoring and restriction of primary to secondary system

leakage through the steam generators. A conservative integrated

leakage program employed by APS provides reasonable assurance than

an orderly unit shutdown will [[Page 27337]] occur prior to a

significant increase in leakage due to failure of a sleeved or

unsleeved tube. The minimal reduction in reactor coolant system

flow, due to sleeving, is considered to have an insignificant impact

on steam generator operation during normal operation or accident

conditions and is bounded by tube plugging evaluations. Therefore,

this change does not involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the licensee's analysis and, based on

that review, it appears that the three standards of Sec. 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket

Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and

2, Rock Island County, Illinois

Date of application for amendment request: February 16, 1993, as

supplemented by letter dated May 2, 1995.

Description of amendment request: As a result of findings by a

Diagnostic Evaluation Team inspection performed by the NRC staff at the

Dresden Nuclear Power Station in 1987, Commonwealth Edison Company

(ComEd, the licensee) made a decision that both the Dresden Nuclear

Power Station and sister site Quad Cities Nuclear Power Station, needed

attention focused on the existing custom Technical Specifications (TS)

used.

The licensee made the decision to initiate a Technical

Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.

The licensee evaluated the current TS for both Dresden and Quad Cities

against the Standard Technical Specifications (STS) contained in NUREG-

0123, ``Standard Technical Specifications General Electric Plants BWR/

4.'' The licensee's evaluation identified numerous potential

improvements such as clarifying requirements, changing TS to make them

more understandable and to eliminate interpretation, and deleting

requirements that are no longer considered current with industry

practice. As a result of the evaluation, ComEd has elected to upgrade

both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.

The TSUP for Dresden and Quad Cities is not a complete adaption of

the STS. The TSUP focuses on (1) integrating additional information

such as equipment operability requirements during shutdown conditions,

(2) clarifying requirements such as limiting conditions for operations

and action statements utilizing STS terminology, (3) deleting

superseded requirements and modifications to the TS based on the

licensee's responses to Generic Letters (GL), and (4) relocating

specific items to more appropriate TS locations.

The February 16, 1993, and May 2, 1995, applications proposed to

upgrade only Section 3/4.10 (Refueling Operations) of the Dresden and

Quad Cities TS.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis.

Implementation of these changes will provide increased reliability

of equipment assumed to operate in the current safety analysis, or

provide continued assurance that specified parameters remain within

their acceptance limits, and as such, will not significantly

increase the probability or consequences of a previously evaluated

accident.

Some of the proposed changes represent minor curtailments of the

current requirements which are based on generic guidance or

previously approved provisions for other stations. The proposed

amendment for Dresden and Quad Cities Station's Technical

Specification Section 3/4.10 are based on STS guidelines or later

operating BWR plant's NRC accepted changes. Any deviations from STS

requirements do not significantly increase the probability or

consequences of any previously evaluated accidents for Dresden or

Quad Cities Stations. The proposed amendment is consistent with the

current safety analyses and has been previously determined to

represent sufficient requirements for the assurance and reliability

of equipment assumed to operate in the safety analysis, or provide

continued assurance that specified parameters remain within their

acceptance limits. As such, these changes will not significantly

increase the probability or consequences of a previously evaluated

accident.

The associated systems that make up the Refueling Systems are

not assumed in any safety analysis to initiate any accident sequence

for Dresden or Quad Cities Stations; therefore, the probability of

any accident previously evaluated is not increased by the proposed

amendment. In addition, the proposed surveillance requirements for

the proposed amendments to these systems are generally more

prescriptive than the current requirements specified within the

Technical Specifications. The additional surveillance requirements

improve the reliability and availability of all affected systems and

therefore, reduce the consequences of any accident previously

evaluated as the probability of the systems outlined within Section

3/4.10 of the proposed Technical Specifications, performing its

intended function is increased by the additional surveillances.

Create the possibility of a new or different kind of accident

from any previously evaluated because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis. Others

represent minor curtailments of the current requirements which are

based on generic guidance or previously approved provisions for

other stations. These changes do not involve revisions to the design

of the station. Some of the changes may involve revision in the

operation of the station; however, these provide additional

restrictions which are in accordance with the current safety

analysis, or are to provide for additional testing or surveillances

which will not introduce new failure mechanisms beyond those already

considered in the current safety analyses.

The proposed amendment for Dresden and Quad Cities Station's

Technical Specification Section 3/4.10 is based on STS guidelines or

later operating BWR plants' NRC accepted changes. The proposed

amendment has been reviewed for acceptability at the Dresden and

Quad Cities Nuclear Power Stations considering similarity of system

or component design versus the STS or later operating BWRs. Any

deviations from STS requirements do not create the possibility of a

new or different kind of accident previously evaluated for Dresden

or Quad Cities Stations. No new modes of operation are introduced by

the proposed changes, considering the acceptable operational modes

in present specifications, the STS, or later operating BWRs.

Surveillance requirements are changed to reflect improvements in

technique, frequency of performance or operating experience at later

plants. Proposed changes to action statements in many places add

requirements that are not in the present technical specifications or

adopt requirements that have been used successfully at other

operating BWRs with designs similar to Dresden and Quad Cities. The

proposed changes maintain at least the present level of operability.

Therefore, the proposed changes do not create the possibility of a

new or different kind of accident from any previously evaluated.

The associated systems that make up the Refueling Systems are

not assumed in any [[Page 27338]] safety analysis to initiate any

accident sequence for Dresden or Quad Cities Stations. In addition,

the proposed surveillance requirements for affected systems

associated with the Refueling Systems are generally more

prescriptive than the current requirements specified within the

Technical Specifications; therefore, the proposed changes do not

create the possibility of a new or different kind of accident from

any previously evaluated.

Involve a significant reduction in the margin of safety because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis. Others

represent minor curtailments of the current requirements which are

based on generic guidance or previously approved provisions for

other stations. Some of the later individual items may introduce

minor reductions in the margin of safety when compared to the

current requirements. However, other individual changes are the

adoption of new requirements which will provide significant

enhancement of the reliability of the equipment assumed to operate

in the safety analysis, or provide enhanced assurance that specified

parameters remain with their acceptance limits. These enhancements

compensate for the individual minor reductions, such that taken

together, the proposed changes will not significantly reduce the

margin of safety.

The proposed amendment to Technical Specification Section 3/4.10

implements present requirements, or the intent of present

requirements in accordance with the guidelines set forth in the STS.

Any deviations from STS requirements do not significantly reduce the

margin of safety for Dresden or Quad Cities Stations. The proposed

changes are intended to improve readability, usability, and the

understanding of technical specification requirements while

maintaining acceptable levels of safe operation. The proposed

changes have been evaluated and found to be acceptable for use at

Dresden and Quad Cities based on system design, safety analysis

requirements and operational performance. Since the proposed changes

are based on NRC accepted provisions at other operating plants that

are applicable at Dresden and Quad Cities and maintain necessary

levels of system, component or parameter (reliability), the proposed

changes do not involve a significant reduction in the margin of

safety.

The proposed amendment for Dresden and Quad Cities Stations will

not reduce the availability of systems associated with the Refueling

Systems when required to mitigate accident conditions; therefore,

the proposed changes do not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: For Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: September 19, 1994, as supplemented by

letter dated April 26, 1995.

Description of amendment request: The amendments would change the

Technical Specifications (TS) to increase the enrichment limits for

fuel stored in the fuel pools and establish restricted loading patterns

and associated burnup criteria for qualifying fuel in the spent fuel

pools. In addition, several administrative changes have been included

in order to provide clarity to the TS and bring them more in line with

the Standard Technical Specifications format. These changes are as

follows:

(1) The TS index is changed to add TS 3/4.9.12 and 3/4.9.13, Tables

3.9-1 and 3.9-2 and Figure 3.9-1.

(2) TS 3/4.9.12, Spent Fuel Pool (SFP) Boron Concentration, is

added to establish a boron concentration limit and to establish a

Limiting Condition for Operation (LCO) for all modes of operation and

to allow the numerical value of the limit to be specified in the Core

Operating Limits Report (COLR).

(3) TS 3/4.9.13, Tables 3.9-1 and 3.9-2 and Figure 3.9-1 are being

added to establish restricted loading patterns for spent fuel storage

and associated burnup criteria.

(4) Corresponding BASES for TSs 3/4.9.12 and 3/4.9.13 are added to

explain the basis for each LCO, Action Statement, and Surveillance

Requirement covered by the subject TSs.

(5) TS 5.6, Fuel Storage, is changed to reflect limits for

criticality analysis for fuel storage.

(6) TS 6.9, Reporting Requirements, is changed to reflect the

inclusion of the SFP boron concentration limit values in the COLR as

established by TS 3/4.9.12.

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.

There is no increase in the probability or consequences of an

accident in the new fuel vault since the only credible accidents for

this area are criticality accidents and it has been shown that

calculated, worst case Keff for this area is (less than or

equal to) 0.95 under all conditions.

There is no increase in the probability of a fuel drop accident

in the Spent Fuel Storage Pool since the mass of an assembly will

not be affected by the increase in fuel enrichment. The likelihood

of other accidents, previously evaluated and described in Section

9.1.2 of the FSAR (Final Safety Analysis Report), is also not

affected by the proposed changes. In fact, it could be postulated

that since the increase in fuel enrichment will allow for extended

fuel cycles, there will be a decrease in fuel movement and the

probability of an accident may likewise be decreased. There is also

no increase in the consequences of a fuel drop accident in the Spent

Fuel Pool since the fission product inventory of individual fuel

assemblies will not change significantly as a result of increased

initial enrichment. In addition, no change to safety related systems

is being made.

Therefore, the consequences of a fuel rupture accident remain

unchanged. In addition, it has been shown that Keff is (less

than or equal to) 0.95, under all conditions. Therefore, the

consequences of a criticality accident in the Spent Fuel Pool remain

unchanged as well.

2. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not create the possibility of a new or

different kind of accident since fuel handling accidents (fuel drop

and misplacement) are not new or different kinds of accidents. Fuel

handling accidents are already discussed in the FSAR for fuel with

enrichments up to 4.0 weight %. As described in Section IV.9 of

Attachment IV, additional analyses have been performed for fuel with

enrichment up to 5.00 weight %. Worst case misloading accidents

associated with the new loading patterns were evaluated. It was

shown that the negative reactivity provided by soluble boron

maintains Keff (less than or equal to) 0.95.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The proposed change does not involve a significant reduction in

the margin of safety since, in all cases, a Keff [less than or

equal to] 0.95 is being maintained. Criticality analyses have been

performed which show that the new fuel storage vault will remain

subcritical under a variety of moderation conditions, from fully

flooded to optimum moderation. As discussed above, the Spent Fuel

Pool will remain sufficiently subcritical during any fuel

misplacement accident.

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 27339]] satisfied. Therefore, the NRC staff proposes to

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730.

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: March 30, 1995, and supplemented May 5,

1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) by separation

of the 24-hour emergency diesel generator (EDG) run from the hot

restart EDG test.

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) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed TS changes would revise the EDG surveillance

criteria to allow the EDG hot-start test with full ESF load

acceptance to be performed separately and independently from the 24-

hour EDG run. The proposed SRs (surveillance requirements) would

continue to demonstrate that the objectives of these two tests are

met. Specifically, the EDGs are shown to be: (1) Capable of starting

and running continuously at full load capability for an interval not

less than 24 hours, and (2) capable of restarting from a full load

temperature condition. The proposed changes would not affect the

EDGs' ability to support mitigation of the consequences of any

previously evaluated accident. Additionally, the proposed changes to

the SRs do not affect the initiating assumptions or progression of

any accident sequence.

Therefore, operation of the facility would not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed TS SR changes do not require any physical changes

to the plant or equipment, and do not impact any design or

functional requirements of the EDGs. The proposed changes do not

create any plant configurations which are prohibited by the TS. The

proposed changes continue to meet the EDG test objectives associated

with demonstrating EDG operability.

Therefore, operation of the facility in accordance with the

proposed amendments would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed TS SR changes do not require any physical changes

to the plant or equipment and do not impact any design or functional

requirements of the EDGs. Surveillance testing in accordance with

the proposed TS will continue to demonstrate the ability of the EDGs

to perform their intended function of providing electrical power to

mitigate design basis transients, consistent with the plant safety

analyses.

Therefore, operation of the facility in accordance with the

proposed amendments would not involve a reduction in a margin of

safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of Sec. 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: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Project Director: David B. Matthews.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: April 7, 1995.

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to relocate the axial power

distribution limits to the Core Operating Limits Report (COLR).

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 relocates the cycle-specific Axial Power

Distribution (APD) limits contained in Figure 1-2 of the Technical

Specifications (TS), to the Core Operating Limits Report (COLR).

This change is consistent with the NRC recommendations of Generic

Letter 88-16, and will not modify the methodology used in generating

the limits nor the manner in which they are implemented. The

methodology used to determine the APD limits is reviewed and

approved by the NRC in accordance with TS 5.9.5. The APD limits will

continue to be determined by analyzing the same postulated events as

previously analyzed. The plant will continue to operate within the

limits specified in the COLR and will take the same remedial actions

if the APD limit is exceeded as required by the current TS.

Therefore, the proposed change would 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 accident previously

evaluated.

There will be no physical alterations to the plant

configuration, changes to setpoint values, or changes to the

implementation of setpoints or limits as a result of this proposed

change. The proposed change only relocates the APD figure from the

TS to the COLR consistent with NRC Generic Letter 88-16. Therefore,

the proposed change does not create the possibility of a new or

different kind of accident from any previously evaluated.

(3) The proposed change does not involve a significant reduction

in a margin of safety.

As indicated above, the implementation of the APD into the COLR,

consistent with the guidance of NRC Generic Letter 88-16, makes use

of the existing safety analysis methodologies and the resulting

limits and setpoints for plant operation. Additionally, the safety

analysis acceptance criteria for operations with the proposed change

have not changed from that use in the current reload analysis.

Therefore, the margin of safety is not reduced due to the relocation

of the APD from the TS and implementation in the COLR.

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 Bateman. [[Page 27340]]

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of amendment requests: April 19, 1995 (Reference LAR 95-03).

Description of amendment requests: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Nuclear Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.8.1.1, ``A.C.

Sources, Operating.'' The specific TS changes proposed are as follows:

(1) TS 4.8.1.1.2b.8), emergency diesel generator (EDG) 24-hour load

run and hot restart surveillance, would be revised to delete the

requirement to perform TS 4.8.1.1.2b.5)b), loss of offsite power (LOOP)

load sequencing surveillance within 5 minutes following the 24-hour

test.

(2) New TS 4.8.1.1.2e. would be added to perform an EDG hot restart

test within 5 minutes of shutting down the EDG after the EDG has

operated for at least 2 hours at a load of greater than or equal to

2484 kW.

(3) TS 4.8.1.1.2b.8), TS 4.8.1.1.2e., and footnote ``*'' on page 3/

4 8-5 would be changed to be cycle-specific with the new TS

requirements effective for Units 1 and 2, Cycle 8 and after.

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.

Demonstrating emergency diesel generator (EDG) hot restart

capability without sequencing loss of offsite power (LOOP) loads

does not invalidate or reduce the effectiveness of the hot restart

test, since normal operating temperatures are achieved prior to the

hot restart test. Sequencing the LOOP loads does not contribute to

verifying that the EDG will start from normal operating

temperatures. The proposed TS 4.8.1.1.[2]e may be performed in any

plant condition since performance of this new surveillance will have

no adverse effect on plant operations. The reliability of the EDGs

is not affected by the proposed changes.

Therefore, the proposed changes do 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 kind of accident from any accident previously

evaluated.

The proposed changes do not involve any physical alterations to

the plant. The proposed changes will not have any adverse effect on

the ability of the EDGs to perform their required safety function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes will not alter any accident analysis

assumptions, initial conditions, or results. Consequently, the

proposed changes do not have any effect on the margin of safety. The

proposed changes to the surveillance requirements would continue to

demonstrate the ability of the EDGs to perform their intended safety

function.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of Sec. 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, PO Box 7442, San Francisco, California 94120.

NRC Project Director: William H. Bateman.

Philadelphia Electric Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric Company,

Docket No. 50-278, Peach Bottom Atomic Power Station, Unit No. 3, York

County, Pennsylvania

Date of application for amendment: November 21, 1994.

Description of amendment request: The proposed change would extend

the Type A test (i.e., Containment Integrated Leak Rate Test (CILRT))

interval on a one-time basis.

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 Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The accidents which are potentially adversely impacted by the

proposed change are any Loss of Coolant Accident (LOCA) inside

primary containment as described in the PBAPS, Units 2 and 3 UFSAR.

The proposed change increases the surveillance interval of the

10 CFR part 50, appendix J Type A test (i.e., Containment Integrated

Leakage Rate Test (CILRT)) from 46 months to 70 months. This test is

performed to determine that the total leakage from containment does

not exceed the maximum allowable primary containment leakage rate

(i.e., designated La) at a calculated peak containment internal

pressure (Pa), as defined in 10 CFR part 50, appendix J. The primary

containment limits the leakage of radioactive material during and

following design bases accidents in order to comply with the offsite

does limits specified in 10 CFR part 100. Accordingly, the primary

containment is not an accident initiator. It is an accident

mitigator. No physical or operational changes to the containment

structure, plant systems, or components would be made as a result of

the proposed change. Therefore, the probability of occurrence of an

accident previously evaluated is not increased.

The failure effects that are potentially created by the proposed

one-time TS change have been considered. The relevant components

important to safety which are potentially affected are the

containment structure, plant systems, and containment penetrations.

There are no physical or operational changes to any plant equipment

associated with the proposed TS change. Therefore, the probability

or consequences of a malfunction of equipment important to safety is

not increased.

The proposed change introduces the possibility that primary

containment leakage in excess of the allowable value (i.e., La)

would remain undetected during the proposed 24 month extension of

the interval between the Type A tests. The types of mechanisms which

would cause degradation of the primary containment can be

categorized into two types. These are: (1) Degradation due to work

which is performed as part of a modification or maintenance activity

on a component or system (i.e., activity-based), or; (2) degradation

resulting from a time-based failure mechanism.

A review of the history of the PBAPS, Unit 3 CILRT results was

performed to evaluate the risk of activity-based and time-based

degradation. This review has determined that the potential for a

time-based and activity-based failure is minimal. The PBAPS LLRT

program would identify most types of penetration leakage. The LLRT

program involves measurement of leakage from Type B and Type C

primary containment penetrations as defined in 10 CFR part 50,

appendix J.

The 10 CFR part 50, appendix J, Type B tests are intended to

detect local leaks and to measure leakage across pressure containing

or leakage-limiting boundaries other than values, such as

containment penetrations incorporating resilient seals, gaskets,

expansion bellows, flexible seal assemblies, door operating

mechanism penetrations that are part of the containment system,

doors, and hatches. 10 CFR part 50, appendix J, Type C testing is

intended to measure reactor system primary containment isolation

valve leakage rates. The frequency of the Type B and Type C testing

is not being altered by the [[Page 27341]] proposed TS change. The

acceptance criterion for Type B and Type C leakage is 0.6 La (i.e.,

0.3% wt/day) which, when compared to the Type A test acceptance

criterion of 0.75 La (i.e., 0.375% wt/day), is a significant portion

of the Type A test allowable leakage.

The proposed TS change only extends the interval between two

consecutive Type A tests. The Type B and Type C tests will be

performed as required. The Type B and Type C tests will continue to

be used to confirm that the containment isolation valves and

penetrations have not degraded. Containment system components that

would not be tested are the containment structure itself and small-

diameter instrumentation lines. Time-based degradation of any of the

instrumentation lines would not likely be identified by faulty

instrument indication or during instrument calibrations that will be

performed during the PBAPS, Unit 3 refueling outage 10. In examining

the potential for a time-based failure mechanism that could cause

significant degradation of the containment structure, we concluded

that the risk, if any, of such a mechanism is small since the design

requirements and fabrication specifications established for the

containment structure are in themselves adequate to ensure

containment leak tight integrity.

Based on the above evaluation, we have concluded that the

proposed TS change will have a negligible impact on the consequences

of any accident previously evaluated.

Although this review concluded that the risk of undetected

primary containment degradation is not increased, the Individual

Plan Examination (IPE) for PBAPS, Units 2 and 3, was also reviewed

in order to access the impact of exceeding the primary containment

allowable leakage rate, if a non-mechanistic activity type (i.e.,

time-based) failure were to occur. The IPE included an evaluation of

the effect of various containment leakage sizes under different

scenarios. The IPE results showed that a containment leakage rate of

35% wt/day would represent less than a 5% increase in risk to the

public of being exposed to radiation. This evaluation was based on a

study performed by Oak Ridge National Laboratory for light water

reactors that evaluated the impact of leakage rates on public risk.

As stated earlier, the current value of La for PBAPS, Unit 3, is

0.5% wt/day, which is significantly less than the 35% wt/day

discussed in the IPE evaluation.

Therefore, the proposed TS change involving a one-time extension

of the Type A test interval and performing the Type A test after the

second appendix J 10-year service period will not involve an

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed TS change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed change is an increase of a surveillance test

interval and does not make any physical or operational changes to

existing plant systems or components. Primary containment acts as an

accident mitigator not initiator. Therefore, the possibility of a

different type of accident than any previously evaluated or the

possibility of a different type of equipment malfunction is not

introduced.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The total primary containment leakage rate ensures that the

total containment leakage volume will not exceed the value assumed

in the safety analyses at the peak accident pressure. As an added

conservatism, the measured overall leakage rate is further limited

to less than or equal to 0.75 La during performance of periodic

tests to account for possible degradation of the containment leakage

barriers between leakage tests. There is the potential that

containment degradation could remain undetected during the proposed

24 month surveillance interval extension and result in the

containment leakage exceeding this allowable value assumed in safety

analysis. A review of the history of the PBAPS, Unit 3 CILRT results

was performed to evaluate the risk of activity-based and time-based

degradation. This review has determined that the potential for a

time-based and activity-based failure is minimal. The PBAPS LLRT

program would identify most types of penetration leakage. The LLRT

program involves measurement of leakage from Type B and Type C

primary containment penetrations as defined in 10 CFR part 50,

appendix J.

The 10 CFR part 50, appendix J, Type B tests are intended to

detect local leaks and to measure leakage across pressure containing

or leakage-limiting boundaries other than valves, such as

containment penetrations incorporating resilient seals, gaskets,

expansion bellows, flexible seal assemblies, door operating

mechanism penetrations that are part of the containment system,

doors, and hatches. 10 CFR part 50, appendix J, Type C testing is

intended to measure reactor system primary containment isolation

valve leakage rates. The frequency of the Type B and Type C testing

is not being altered by the proposed TS change.

Therefore, we have concluded that the proposed extended test

interval would not result in a non-detectable PBAPS, Unit 3 primary

containment leakage rate in excess of the allowable value (i.e.,

0.5% wt/day) established by the TS and 10 CFR part 50, appendix J.

Therefore, the proposed TS change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: April 4, 1995.

Description of amendment request: The amendment would provide a

one-time interval extension for the Type A test required by 10 CFR part

50, appendix J. The extension would allow the test to be conducted

during the thirteenth refueling outage, rather than the twelfth

refueling outage.

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. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change will provide a one-time exemption from 10

CFR part 50, appendix J Section III.D.1(a) leak rate test schedule

requirement. This change will allow for a one-time test interval for

Type A Integrated Leak Rate Tests (ILRTs) of 65+/-10 months.

Leak rate testing is not an initiating event in any accident,

therefore, this proposed change does not involve a significant

increase in the probability of a previously evaluated accident.

Type A tests are capable of detecting both local leak paths and

gross containment failure paths. The history at Salem Generating

Station Unit 1 (SGS1) demonstrates that Type B and C Local Leak Rate

Tests (LLRTs) have consistently detected any excessive local

leakages. SGS1 has passed all of its ILRTs with significant margin.

Administrtive controls govern the maintenance and testing of

containment penetrations such that the probability of excessive

penetration leakage due to improper maintenance or valve

misalignment is very low. Following any maintenance that could

affect the leakage characteristics of any containment penetration,

an LLRT is performed to ensure acceptable leakage levels. Following

any LLRT on a containment isolation valve, an independent valve

alignment check is performed before declaring the penetration

OPERABLE. Therefore, Type A testing is not necessary to ensure

acceptable leakage rates through containment penetrations.

While Type A testing is not necessary to ensure acceptable

leakage rates through [[Page 27342]] containment penetrations, Type

A testing is necessary to demonstrate that there are no gross

containment failures. Structural failure of the containment is

considered to be a very unlikely event, and in fact, since SGS1 has

been in operation, it has never failed a Type A ILRT. Therefore, a

one-time exemption increasing the interval for performing an ILRT

does not result in a significant decrease in the confidence in the

leak tightness of the containment structure.

Therefore, this proposed change does not result in a significant

increase of the probability or consequences of any previously

evaluated accident.

2. Will not create the possibility of a new or different kind of

accident from any accident previously evaluated.

This proposed change allows a one-time interval of 65+/-10

months for the next ILRT. The method of performing the test is not

changed. No new accident modes are created by extending the testing

intervals. No safety-related equipment or safety functions are

altered as a result of this change. A one-time extension of the ILRT

test interval has no influence on, nor does it contribute in any way

to, the possibility of a new or different kind of accident or

malfunction from those previously analyzed.

3. Will not involve a significant reduction in a margin of

safety.

The purpose of the existing schedule of ILRTs is to ensure that

the release of radioactive materials will be restricted to those

leak paths and leak rates assumed in accident analyses. The relaxed

schedule for ILRTs does not allow for relaxation of Type B and C

LLRTs. Therefore, methods for detecting local containment leak paths

and leak rates are unaffected by this proposed change. Given that

the test history for ILRTs shows no failure during plant life, a

one-time increase of the test interval does not lead to a

significant probability of creating a new leakage path or increased

leakage rates, and the margin of safety inherent in existing

accident analyses is maintained. Therefore, this change does not

involve a significant reduction kin the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey 08079.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington, DC 20005-3502.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: May 4, 1995.

Description of amendment request: The amendment would authorize a

one-time extension for the Type A test (overall integrated containment

leakage rate) that is required by 10 CFR part 50, appendix J. The

current Technical Specification would require that this test be

conducted by July 7, 1995. The amendment would allow this test to be

conducted by November 30, 1995.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change involves no hardware changes, no changes to

the operation of any systems or components, and no changes to

existing structures. This change is temporary, allowing a one-time

extension of a specific surveillance requirement for cycle 12 to

allow surveillance testing to coincide with the twelfth refueling

outage. The proposed surveillance interval extension is short and

will not result in any significant reduction in structural

reliability nor will the extension affect the ability of the

structure in performing its intended functions. to preclude the

possibility of an undetected containment failure/leakage at a valve

or penetration seal, Type ``B'' and ``C'' tests will continue to be

performed as required by the Technical Specifications. Therefore,

this change will not involve a significant increase in the

probability or consequences of any accidents previously evaluated.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

Extending the surveillance interval for the performance of

specific testing will not create the possibility of any new or

different kinds of accident. No changes are required to any system

configurations, plant equipment, or analyses. Therefore, this change

will not create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. Will not involve a significant reduction in a margin of

safety.

The proposed change will not alter any assumptions, initial

conditions, or results of any accident analyses. The safety limits

assumed in the accident analyses and the design function of the

structure required to mitigate the consequences of any postulated

accidents will not be changed since only the surveillance interval

is being extended. Historical performance indicates a high degree of

reliability, and surveillance testing performed during continued

plant operation will verify that Salem 1 will remain within analyzed

limits. Consequently, the change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of amendment request: April 18, 1995.

Description of amendment request: The amendments would delete the

quarterly leak rate test for the containment pressure-vacuum relief

valves which is presently required because of the valves' resilient

seat material. The resilient valve seat material will be replaced with

a hard seat (metal to metal) design. The valves would still remain in

the 10 CFR part 50 appendix J, Type C leak rate test program.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The containment pressure/vacuum relief valves are normally

closed, and are used under administrative control to maintain

containment internal pressure within -1.5 psig and +0.3 psig, as

required by SGS Technical Specifications. The pressure/vacuum relief

valves are relied upon for containment isolation and automatically

close on high containment pressure or high containment atmosphere

radioactivity. The pressure/vacuum relief system does not affect the

probability of any previously evaluated accident.

The containment isolation function of the pressure/vacuum relief

valves limits the consequences of a radiological release inside

containment (i.e., Loss of Coolant Accident). The proposed changes

to eliminate quarterly pressure drop (leak rate) testing would not

increase the consequences of any previously evaluated accident. The

valve flow characteristics and closure time requirements are not

affected. The valves will continue to be subject to the Type C leak

rate test criteria of 10 CFR part 50, appendix J. The deletion of

the augmented quarterly test requirement is justified by replacement

of the resilient [[Page 27343]] valve seat material (which has a

history of degradation and loss of leaktightness) with a metal to

metal seating design.

2. Do not create the possibility of a new or different kind of

accident from any accident previously evaluated.

Eliminating quarterly leak rate testing based on improved valve

design would not result in any new or different kind of accident.

The valves would continue to perform the containment isolation

function consistent with the plant safety analyses, and would not

adversely affect the initiation or progression of any accident

sequence.

(3) Do not involve a significant reduction in a margin of

safety.

This proposal involves replacement of the existing pressure/

vacuum relief valves, which have resilient seating material, with

valves using a hard seat (metal to metal design). Based on the

improved design and operating experience of the replacement valves,

augmented quarterly leak rate testing is no longer necessary or

appropriate to verify leaktightness of the valves. Periodic leak

rate testing will continue to be performed in accordance with 10 CFR

part 450, appendix J. The pressure/vacuum relief valves will

continue to maintain their containment isolation capability such

that no margin of safety is affected by the proposed changes.

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: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plants, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: May 3, 1995 (TS 93-09).

Descripton of amendment request: The proposed change would revise

the implementation schedule for Amendment Nos. 182 and 174 from that

stated in the amendments when they were approved by the Commission by

letter dated May 24, 1994. As issued, the amendments reflected the

licensee's plans to implement the changes for both units during the

Unit 2 Cycle 6 refueling outage. However, by letter dated August 19,

1994, the licensee requested that implementation be delayed to 1995.

This request was granted by Amendment Nos. 188 and 180 for Units 1 and

2 respectively by letter dated October 17, 1994. By letter dated May 3,

1995, the licensee informed the staff that evaluation of the design

changes have concluded that significant safety risks would be involved

with modification activities associated with installation. Therefore,

the licensee has requested that implementation of the amendment be

changed to specify that the amendment will be implemented along with

the related plant modifications, without specifying the date when the

modifications would be performed. No changes to the technical

specification pages other than those approved when the amendments were

issued are needed.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

determined that the no significant hazards consideration exists. This

analysis was provided in the original submittal for the amendment from

the licensee dated October 1, 1993, and was used in the preparation of

the amendments. The licensee has determined that this analysis remains

valid for the proposed revision and that the changes do not constitute

a significant hazard. The staff previously issued the proposed finding

in the Federal Register (59 FR 4947 and 59 FR 47182) and there were no

public comments on the finding. This analysis is reproduced as follows:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed revision supports the implementation of design

logic and setpoint changes to the loss-of-power relaying. This

relaying is designed to ensure adequate voltage is available to

safety-related loads in order to enhance their operability and

support accident mitigation functions and to provide for auxiliary

feedwater (AFW) pump starts. The design changes alter relay logic

and delete unnecessary relaying, but do not change the diesel

generator (D/G) start and load-shedding actuations that result from

loss-of-power conditions. Therefore, no new actuations or functions

have been created; and because the existing and proposed functions

provide for accident mitigation considerations that are not the

source of an accident, the probability of an accident is not

increased. The deletion of the 6.9-kilovolt shutdown board normal-

feeder undervoltage relays actually reduces the potential for

inadvertent shutdown board blackouts as a result of short-duration

voltage transients or instrument failures.

The setpoints and time delays for loss-of-power functions have

been modified based on the guidelines developed by the Electrical

Distribution System Clearinghouse as evaluated and determined

through detailed analysis by TVA. This design is documented in TVA

Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is

available for NRC review at the SQN site. The assigned values are

conservative settings that will ensure adequate voltage is supplied

to safety-related loads for accident mitigation and safety functions

under normal, degraded, and loss-of-offsite-power voltage conditions

with appropriate time delays to prevent damage to electrical loads

and minimize premature or unnecessary actuations. The identification

of loss-of-voltage conditions is enhanced by the design changes to

ensure the timely sequencing of loads onto the D/G and the

initiation of AFW pump starts for accident mitigation. Because there

are no reductions in safety functions resulting from the design

logic, setpoint, and time-delay changes to the loss-of-power

instrumentation and offsite dose levels for postulated accidents

will not be increased, the consequences of an accident are not

increased.

The applicable mode addition, TS 3.0.4 exclusion deletion, and

response time measurement clarification incorporated in the proposed

change do not affect plant functions. These changes reflect the

requirements that SQN has been maintaining and serve to clarify the

requirements to provide consistency of application and easier

understanding. The AFW footnote addition and bases revision only

clarify operability conditions that are consistent with the plant

design for the AFW pump and loss-of-power instrumentation. Because

there are no changes to plant functions or operations, these

revisions have no impact on accident probabilities or consequences.

2. Create the possibility of a new or different kind of accident

from any previously analyzed.

As described above, the loss-of-power instrumentation ensures

adequate voltage to safety-related loads by initiating D/G starts

and load shedding and provides for AFW pump starting, but is not

considered to be the source of an accident. Although the design

logic, setpoint, and time-delay actuation criteria have changed, the

output functions to various plant systems that actuate for load

shedding and D/G starts remain the same. Therefore, actuation

criteria have been affected, but not safety functions, and the TVA

evaluation has confirmed that the new design enhances the ability to

maintain adequate voltage to support safety functions. Since safety

functions have not changed and the new loss-of-power instrumentation

design continues to support operability of safety-related equipment,

no new or different accident is created.

The applicable mode addition, TS 3.0.4 exclusion deletion, and

response time measurement clarification, as well as the AFW

operability clarifications, do not affect plant functions and will

not create a new accident.

3. Involve a significant reduction in a margin of safety.

The proposed loss-of-power TS changes support design logic,

setpoint, and time-delay requirements that have been verified by

[[Page 27344]] TVA analysis to provide acceptable voltage levels for

safety-related components. In determining the acceptability of these

voltage levels, the minimum voltage for operation as well as

detrimental component heating resulting from sustained degraded-

voltage conditions were considered. This design ensures that safety-

related loads will be available and operable for normal and accident

plant conditions. The applicable mode addition, TS 3.0.4 exclusion

deletion, response time measurement clarification, and AFW

operability clarifications provide enhancements to TS requirements

and do not affect plant functions. Therefore, no safety functions

are reduced by these changes and there is no reduction in the margin

of safety.

The NRC has reviewed the licensee's analysis and, based on this

review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: April 28, 1995.

Description of amendment request: The proposed amendment would

extend for one more operating cycle an exception to Limiting Condition

for Operation (LCO) 3.0.4 as it applies to the Technical Specification

for the main steam isolation valve leakage control system. The existing

LCO 3.0.4 exception was issued by Amendment 63 to the Operating

License, and will expire upon completion of the fifty cycle of plant

operation. The extension is proposed for the duration of the sixth

cycle of operation to permit completion of activities necessary to

implement the most appropriate permanent resolution for the issue of

secondary containment bypass leakage through the main steam line

drains.

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.

This License Amendment application proposes an extension for one

operating cycle of the exception to Limiting Condition for Operation

for Operation (LCO) 3.0.4 as it applies to the Technical

Specification for the MSIV [main steam isolation valve] Leakage

Control system. This extension is proposed for the duration of the

sixth cycle of PNPP (Perry Nuclear Power Plant) operation, to permit

completion of activities necessary to implement the most appropriate

permanent resolution for the issue of secondary containment bypass

leakage through the Main Steam Line drains. During the sixth cycle,

the drains will remain in their current configuration, which seals

off the bypass leakage path. The sealed drain path results in a

temporary inoperability of the Inboard MSIV Leakage control system

(MSIV-LCS) subsystem when the plant is operated below 50% power, due

to condensate build-up in the bottom of the steam lines between the

MSIVs. The requested 3.0.4 exception is necessary to permit plant

startups with this temporary inoperability, for the duration of the

sixth operating cycle.

The probability of occurrence of a previously evaluated accident

is not affected by the proposed extension of the LCO 3.0.4 exception

since no change to the plant or to the manner in which the plant is

operated is involved. The existing plant configuration will be

maintained for another operating cycle, and possible concerns

resulting from that configuration have been analyzed. The extra

weight of the water pooled between the MSIVs was analyzed with

respect to piping supports and seismic considerations and was found

to be acceptable, and any condensate that is carried past the

outboard MSIVs will be drained to the condenser by drain connections

downstream of the outboard MSIVs before it can reach the turbine.

The temporary inoperability of the Inboard MSIV-LCS when below 50%

power has no impact on accident initiation probability, since LCS

does not serve to prevent accidents, but is only used in mitigating

the consequences of Loss of Coolant Accidents that have already

occurred.

The consequences of an accident are not significantly increased

in that the Outboard MSIV-LCS will be available to perform the MSIV-

LCS function by mitigating the consequences of a Loss of Coolant

Accident (LOCA) during the temporary period in which the Inboard

MSIV-LCS is unavailable. Any condensate that is carried past the

outboard MSIVs will be drained to the condenser by drain connections

downstream of the outboard MSIVs; therefore no impairment of the

Outboard MSIV-LCS will result from condensed water.

The Action statement for one inoperable LCS subsystem remains

the same, and the limits plant operation to the previously

established 30-day Allowable Outage Time. The Action required if

both the subsystems of MSIV-LCS were to become inoperable also

remains the same. The MSIV function of isolating the Main Stream

Lines is also unaffected by the existing plant configuration, since

MSIV performance will not be affected by the existence of

accumulated water in the bottom of the steam lines between the MSIVs

during the plant operation below 50% power. Therefore, if necessary,

the Main Steam Lines will be isloated, and leakage past the MSIVs

will be routed for filtration as in the design-basis radiological

analyses, and the consequences of previously evaluated accidents

will remain unaffected.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed change to permit inoperability of the Inboard MSIV-

LCS during periods of startup and power ascension to 50% RTP (rated

thermal power) and during shutdown below 50% RTP does not create the

possibility of a new or different kind of accident from any

previously evaluated. The Inboard MSIV-LCS is only credited during a

Recirculation Line Break LOCA wherein Reactor Coolant System

depressurization occurs. The temporary unavailability of the Inboard

MSIV-LCS. the amendment to the Technical Specifications is an

administrative change that does not involve any change to the

current plant design or methods of operation. No new plant equipment

failure modes or accident initiators are introduced by the extension

of the LCO 3.0.4 exception.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The response to the Recirculation Line Break LOCA will not be

significantly affected since the Outboard MSIV-LCS can be assumed to

be available. Allowing entry into Operational Conditions 1, 2 and 3

while utilizing the existing Action statement does not significantly

reduce the margin of safety since the duration of time allowed for

remaining in that Action statement is not increased. The proposed

change will have no adverse impact on the reactor coolant system

pressure boundary nor will any other system protective boundary or

safety limit be affected.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Gail H. Marcus. [[Page 27345]]

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: May 1, 1995.

Description of amendment request: The proposed amendment would

eliminate selected response time testing requirements, and incorporate

guidance provided by Generic Letter 93-08, ``Relocation of Technical

Specification Tables of Instrument Response Time Limits.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

For those proposed changes dealing with the elimination of

selected response time test requirements, the purpose of the

proposed Technical Specification change is to eliminate response

time testing requirements for selected components in the Reactor

Protection System, Isolation system, and Emergency Core Cooling

System. The BWR Owners' Group has completed an evaluation which

demonstrates that the response time testing is redundant to other

Technical Specification required testing. These other tests, in

conjunction with actions taken in response to NRC Bulletin 90-01,

``Loss of Fill-Oil in Transmitters Manufactured by Rosemount,'' and

Supplement 1, are sufficient to identify failure modes or

degradations in instrument response time and ensure operation of the

associated systems within acceptable limits. There are no known

failure modes that can be detected by response time testing that

cannot also be detected by the other required Technical

Specification testing. This evaluation was documented in NEDO-32291,

``System Analyses for Elimination of Selected Response Time Testing

Requirements,'' January 1994, and the letter from T. Green to P.

Loeser dated April 15, 1994 which were approved by an NRC Safety

Evaluation dated December 28, 1994. The applicability of this

evaluation to the Perry Nuclear Power Plant (PNPP) has been

confirmed. In addition, PNPP will complete the additional actions

identified in the NRC staff's Safety Evaluation of NEDO-32291.

Because of the continued application of other existing Technical

Specification required tests such as channel calibrations, channel

checks, channel functional tests, and logic system functional tests,

the response times of these systems will be maintained within the

acceptance limits assumed in plant safety analysis and required for

successful mitigation of an initiating event. The proposed Technical

Specification changes do not affect the capability of the associated

systems to perform their intended function within their required

response time, nor do the proposed changes themselves affect the

operation of any equipment. As a result the proposed changes dealing

with elimination of selected response time tests do not involve a

significant increase in the probability or the consequences of an

accident previously evaluated.

For those changes dealing with moving the surveillance

requirement for ECCS RESPONSE TIME testing from the instrumentation

section to the system section of the Technical Specifications, no

change in testing requirements (other than the elimination of the

instrument loops implemented as part of the NEDO-32291 changes) has

been introduced. The relaxation in Applicability does not increase

the probability or the consequences of an accident previously

evaluated, since there are no design basis events during OPERATIONAL

CONDITION 4 and 5 where ECCS systems are relied upon.

For those changes dealing with relocation of the response time

limits from Technical Specification Tables and into the Updated

Safety Analysis Report (USAR), the proposed changes are

administrative in nature in that the test requirements and time

limits are still requirements, but the placement of the limits have

been relocated from the Technical Specifications and into the USAR.

Therefore these changes do not involve a significant increase in the

probability or the consequences of an accident previously evaluated.

2. The changes do not create the possibility of a new or

different kind of accident from any previously evaluated.

None of the proposed Technical Specification changes affect the

capability of the associated systems to perform their intended

function within the acceptance limits assumed in plant safety

analyses and required for successful mitigation of an initiating

event. The proposed changes also do not change the manner in which

any plant equipment is operated. Therefore, the proposed changes do

not create the possibility of a new or different kind of accident

from any previously evaluated.

3. The changes do not involve a significant reduction in the

margin of safety.

The current Technical Specification response times are based on

the maximum allowable value assumed in the plant safety analyses.

These analyses conservatively establish the margin of safety. As

described above, the proposed Technical Specification changes do not

affect the capability of the associated systems to perform their

intended function within the allowed response time used as the basis

for the plant safety analyses. Plant and system response to an

initiating event will remain in compliance within the assumptions of

the safety analyses, and therefore the margin of safety is not

affected.

Although not explicitly evaluated, the proposed Technical

Specification changes dealing with response time testing elimination

will provide an improvement to plant safety and operation by

reducing the time safety systems are unavailable, reducing safety

system actuation, reducing plant shutdown risk, limiting radiation

exposure to plant personnel, and eliminating the diversion of key

personnel to conduct unnecessary testing. Therefore, the proposed

changes do not result in a significant reduction in a margin of

safety, and may result in an overall increase in the margin of

safety.

The changes dealing with relocation of the time response limits

from the Technical Specifications to the USAR is an administrative

change that does not affect either the requirements to perform

response time testing or the limits associated with the response

time tests. Future changes to the limits will be controlled by 10

CFR 50.59. Therefore, this portion of the change does not result in

a significant decrease 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: April 26, 1995.

Description of amendment request: the proposed amendment would

revise Technical Specification (TS) Surveillance Requirements 3/4.7.6

and associated Bases to reduce the upper limit on the control room

filtration subsystem flow rate. It would also adopt ASTM D-3803-1989 as

the laboratory testing standard for control room filtration and control

building pressurization charcoal absorber.

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 revision does not involve a significant hazards

consideration because operation of Callaway Plant with this change

would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Overall protection system performance will remain within the

bounds of the accident analysis documented in FSAR Chapter 15 * * *

since no hardware changes are proposed. [[Page 27346]]

The Control Room Emergency Ventilation System (CREVS) will

continue to function in a manner consistent with the above analysis

assumptions and the plant design basis. There will be no degradation

in the performance of or an increase in the number of challenges to

equipment assumed to function during an accident situation.

These Technical Specification revisions do not involve any

hardware changes nor do they affect the probability of any event

initiators. The change to the control room filtration flow rate is

consistent with the original licensing basis and will ensure an

average atmosphere residence time of greater than or equal to 0.25

sec. There will be no change to ESF (engineered safety feature)

actuation setpoints or accident mitigation capabilities. The

laboratory testing will demonstrate the required absorber

performance after a design basis LOCA (loss-of-coolant accident).

The control room dose analyses assume a total flow rate through

the control room filtration units that is less than the proposed

upper limit. As such, there will be no changes required to the

control room dose analyses.

Based on the above, these Technical Specification changes will

not increase the probability or consequences of an accident or

malfunction.

(2) Create the possibility of a new or different kind of

accident from any accident previously evaluated.

As discussed above, there are no hardware changes associated

with these Technical Specification revisions nor are there any

changes in the method by which any safety-related plant system

performs its safety function.

Revisions to the Surveillance Requirements for the CREVS will

ensure that the control room does analysis assumptions made in

support of OL (operating license) Amendment No. 96 are valid.

Changes to the control room filtration unit flow rate are more

limiting than that currently specified and have already been

implemented by resetting the open limit switches on the respective

units' outlet dampers. This flow rate is consistent with the design

basis for the filtration units as originally licensed.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of these 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 kind of accident is

not created.

(3) Involve a significant reduction in a margin of safety.

There will be no margin reduction since these changes are in the

conservative direction and have already been approved by NRC via the

approval of OL Amendment No. 96. The reduced upper bound flow rate

for the control room filtration units is consistent with their

design basis and will maintain an average atmosphere residence time

greater than or equal to 0.25 sec under both clean and dirty filter

conditions. The new charcoal absorber sample laboratory testing

protocol is more stringent than the current testing practice and

more accurately demonstrates the required performance after a design

basis LOCA.

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. There will no impact on the

overpower limit, DNBR (departure from nucleate boiling ratio)

limits, FQ, F[delta]H, LOCA PCT (peak cladding temperature),

peak local power density, or any other margin of safety. These

changes will ensure that the criteria of GDC 19 are met.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks.

Manitowoc County, Wisconsin

Date of amendment request: April 17, 1995.

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Section 15.6.2, ``Organization,''

and TS Section 15.6.3, ``Facility Staff Qualifications.'' The training

requirements for the Operations Manager and other staff would be

changed to provide staffing flexibility.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated;

create the possibility of a new or different kind of accident from

any previously evaluated; or create the possibility of a new or

different kind of accident from any previously evaluated.

1. The proposed change affects only an administrative control,

which was based on industry guidance in ANSI N18.1-1971, that

recommended the Operations Manager hold an SRO (senior reactor

operator) license. This administrative control is being updated to

meet the current guidance in ANSI/ANS 3.1-1987.

2. The proposed qualification requirements for the Operations

Manager ensures the individual filling the position meets knowledge

levels equivalent to the present requirements. It also ensures that

individuals responsible for directing the activities of licensed

operators continue to hold SRO licenses as required by 10 CFR

50.54(l).

3. Since the proposed specifications ensure regulatory

requirements are met and ensures knowledge levels equivalent to

existing license requirements for operations management, the

proposed changes are considered administrative. The design of plant

systems and equipment is not being altered. Plant operations will

continue to be directed and performed by qualified personnel.

Therefore, the probability or consequences of accidents previously

evaluated are not affected, a new or different type of accident is

not created, nor is a margin of safety reduced.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: April 27, 1995.

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Table 15.3.5-1, ``Engineered Safety

Features Initiation Instrument Setting Limits,'' and TS Table 15.35-3,

``Engineered Safety Features.'' Setting limits would be modified and

references would be changed. The bases for TS Section 15.3.5,

``Instrumentation System,'' would also be changed to be consistent with

the TS changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on

the probability of initiating events for these accidents.

[[Page 27347]] Initiating events for accidents previously evaluated

for Point Beach include: control rod withdrawal and drops, CVCS

(chemical and volume control system) malfunction (Boron Dilution),

startup of an inactive reactor coolant loop, reduction in feedwater

enthalpy, excessive load increase, losses of reactor coolant flow,

loss of external electrical load, loss of normal feedwater, loss of

all AC power to the auxiliaries, turbine overspeed, fuel handling

accidents, accidental releases of water liquid or gas, steam

generator tube rupture, steam pipe rupture, control rod ejection,

and primary coolant system ruptures.

This license amendment request proposes to correct some minor

errors, include appropriate operability requirements for the

modification to include the safety injection signal in the time

delay for the 4.16KV degraded voltage protection logic, slightly

lower the degraded voltage setting limit, change the format of the

4.16 KV degraded voltage and loss of voltage setting limits, and

change the time delays associated with the 4.16 KV degraded voltage,

4.16 KV loss of voltage and 480 V loss of voltage protection

functions.

These proposed changes do not cause an increase in the

probabilities of any accidents previously evaluated because these

changes will not cause an increase in the probability of any

initiating events for accidents previously evaluated. In particular,

these proposed changes affect time delay and format of the setting

limits associated with the 4.16 KV degraded voltage, 4.16 KV loss of

voltage, and 480 V loss of voltage protection functions. These are

protection functions and do not cause accidents.

The consequences of the accidents previously evaluated in the

PBNP FSAR (Final Safety Analysis Report) are determined by the

results of analyses that are based on initial conditions of the

plant, the type of accident, transient response of the plant, and

the operation and failure of equipment and systems. The changes

proposed in this license amendment request provide appropriate

limiting conditions for operation, action settlements, allowable

outage times, setting limits, and time delays for the Point Beach

Nuclear Plant Technical Specifications for the 4.16 KV degraded

voltage, 4.16 KV loss of voltage, and 480 V loss voltage protection

functions.

The proposed changes affect functions that are required to

ensure the proper operation of engineered features equipment. The

proposed changes do not increase the probability of failure of this

equipment or its ability to operate as required for the accidents

previously evaluated in the PBNP FSAR.

The modifications to reduce the time delay limit associated with

the 4.16 KV degraded voltage protection function when the degraded

voltage condition is coincident with a safety injection signal, have

been designed and installed in accordance with the requirements for

PBNP. The probability of occurrence of degraded voltage conditions

at PBNP has not been increased. The modifications and proposed

Technical Specifications will ensure the proper operation of ESF

(engineered safety feature) equipment. These changes do not increase

the possibility of failure of this equipment.

Therefore, this proposed license amendment does not affect the

consequences of any accident previously evaluated in the Point Beach

Nuclear Plant FSAR, because the factors that are used to determine

the consequences of accidents are not being changed.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

New or different kinds of accidents can only be created by new

or different accident initiators or sequences. New and different

types of accidents (different from those that were originally

analyzed for Point Beach) have been evaluated and incorporated into

the licensing basis for Point Beach Nuclear Plant. Examples of

different accidents that have been incorporated into the Point Beach

Licensing basis include anticipated transients without scram and

station blackout.

The changes proposed by this license amendment request do not

create any new or different accident initiators or sequences because

these changes to the 4.16 KV degraded voltage, 4.16 KV loss of

voltage, and 480 V loss of voltage protection functions will not

cause failures of equipment or accident sequences different than the

accidents previously evaluated. Therefore, these modifications and

proposed Technical Specification changes do not create the

possibility of an accident of a different type than any previously

evaluated in the Point Beach FSAR.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design

and operation of the reactor and containment and the safety systems

that provide their protection.

The changes proposed by this license amendment request provide

the appropriate setting limits and time delays for the 4.16 KV

degraded voltage, 4.16 KV loss of voltage, and 480 V loss of voltage

protection functions. This ensures that the safety systems that

protect the reactor and containment will operate as required. The

design and operation of the reactor and containment are not affected

by these proposed changes. Therefore, the margins of safety for

Point Beach are not being reduced because the design and operation

of the reactor and containment are not being changed and the safety

systems that provide their protection that are being changed are

being modified in accordance with the applicable design and

installation requirements for Point Beach Nuclear Plant.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued 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.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: April 21, 1995.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) 3.1.2.4, ``Charging Pumps-

Operating,'' by adding a note that indicates that the provisions of TS

3.0.4 and 4.0.4 are not applicable for entry into MODE 4 from MODE 5.

Date of publication individual notice in Federal Register: May 2,

1995 (60 FR 21558).

Expiration date of individual notice: June 1, 1995.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, Connecticut 06360.

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

[[Page 27348]] 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 Ch. 1, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document rooms for

the particular facilities involved.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of application for amendment: August 19, 1994, as supplemented

November 3, 1994.

Brief description of amendment: The amendment requests a line-item

improvement to the Radiological Effluent Technical Specifications

pursuant to the guidance of Generic Letter 89-01 and incorporates the

requirements of revised 10 CFR part 20 and 10 CFR 50.36a.

Date of issuance: May 1, 1995.

Effective date: May 1, 1994.

Amendment No.: 58.

Facility Operating License No. NPF-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: October 12, 1994 (60 FR

51617) The Commission's related evaluation of the amendment, and NRC's

response to the public comments received, are contained in a Safety

Evaluation dated May 1, 1995.

No significant hazards consideration comments received: Yes.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: November 22, 1994, as

supplemented by letters dated January 30, March 2, March 13, and May 2,

1995.

Brief description of amendments: The amendments revise Technical

Specification 3.8 to establish restricted loading patterns and

associated burnup criteria for placing fuel in the Oconee spent fule

pools. In addition, the Design Features sections associated with the

reactor and fuel storage are also revised.

Date of issuance: May 3, 1995.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: 209, 209, and 206.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 15, 1995 (60

FR 8746); Re-Noticed March 29, 1995 (60 FR 16185).

The May 2, 1995, letter did not change the scope of the November

22, 1994, application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 3, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: April 10, 1995, as supplemented

April 12, 1995, and April 20, 1995.

Brief description of amendment: This amendment revises Technical

Specification 4.6.2.2.d to delete the reference to the specific test

acceptance criteria for the Containment Recirculation Spray Pumps and

replace the specific test acceptance criteria with reference to the

developed head required by the plant's safety analysis. In addition,

the 18-month test frequency would be replaced with the test frequency

requirements specified in the IST Program. The current footnote (1)

pertaining to the performance of recirculation spray pump 2RSS*P21A

would be deleted.

Date of issuance: May 3, 1995.

Effective date: May 3, 1995.

Amendment No.: 68.

Facility Operating License No. NPF-73: Amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes (60 FR 19417, April 18, 1995) That notice provided

an opportunity to submit comments on the Commission's proposed no

significant hazards consideration determination. No comments have been

received. The notice also provided for an opportunity to request a

hearing by May 18, 1995, but indicated that if the Commission makes a

final no significant hazards consideration any such hearing would take

place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of

exigent circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated May 3,

1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas

Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

Date of amendment request: August 30, 1994 as supplemented January

19, 1995.

Brief description of amendments: The amendments changed

requirements related to the site perimeter security system.

Date of issuance: April 28, 1995.

Effective date: April 28, 1995.

Amendment Nos.: Unit 1--Amendment No. 180; Unit 2--Amendment No.

161

Facility Operating License Nos. DPR-51 and NPF-6: Amendments

revised the licenses.

Date of initial notice in Federal Register: April 12, 1995 (60 FR

18625).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 28, 1995.

No significant hazards consideration comments received: No.

[[Page 27349]]

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 14, 1993, as supplemented by

letter dated March 3, 1995.

Brief description of amendment: The amendment changed the Appendix

A Technical Specifications by removing the reactor vessel material

specimen withdrawal schedule and by updating the reactor coolant system

pressure-temperature (P-T) curves.

Date of issuance: May 8, 1995.

Effective date: May 8, 1995.

Amendment No.: 106.

Facility Operating License No. NPF-38.: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 19, 1994 (59 FR

2867).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 8, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: October 20, 1994.

Brief description of amendments: These amendments change the

definition of ``core alteration'' to exclude the movement of items not

associated with reactivity. The second change involves allowing the

personnel airlock (PAL) doors to remain open during fuel movement and

core alterations under certain conditions.

Date of issuance: May 11, 1995.

Effective date: May 11, 1995.

Amendment Nos.: 173 and 167.

Facility Operating License No. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55869).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 11, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: February 28, 1995.

Brief description of amendment: The amendment revises Technical

Specification (TS) Section 6.5.1.12 to delete the requirement to render

determinations in writing with regard to whether or not activities

listed in TS Sections 6.5.1.2 and 6.5.1.5 constitute an unreviewed

safety question. These activities are changes to Appendix A Technical

Specifications (6.5.1.2) and investigations of all violations of the

TSs (6.5.1.5). This change is consistent with NUREG-1433 Standard

Technical Specifications General Electric Plants, BWR/4 Revision 0,

dated September 28, 1992.

Date of issuance: May 1, 1995.

Effective date: May 1, 1995.

Amendment No.: 180.

Facility Operating License No. DPR-16.: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16188).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated May 1, 1995.

No significant hazards consideration comments received: Yes.

By letter dated April 5, 1995, Mr. Kent W. Tosch, of the State of

New Jersey Department of Environmental Protection commented that they

concur with GPU Nuclear's rationale that these unreviewed safety

question reviews serve no value since these activities specifically

require NRC review and approval. The State official had no other

comments.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas, Docket

Dos. 50-498 and 50-499, South Texas Projects, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: February 15, 1995.

Brief description of amendment: The amendment modified Technical

Specification 4.6.2.3.a.2 (and associated Bases) to reflect the reactor

containment fan cooler flow rate assumed in the accident analysis and

to specify that this flow is provided by the component cooling water

system.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995, to be implemented within 30 days.

Amendment Nos.: Unit 1--Amendment No. 74; Unit 2--Amendment No. 63.

Facility Operating License Nos. NPF-76 AND NPF-80. The amendment

revised the Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16189) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

J.M. Hodge Learning Center, 911 Boling Highway, Wharton, TX 77488.

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 10, 1995.

Brief description of amendment: The amendment changes Technical

Specification 3.3.2.1, ``Control Rod Block Instrumentation,'' to revise

two surveillance requirements and their associated notes for the Rod

Withdrawal Limiter mode of the Rod Pattern Control System. The changes

are consistent with the Clinton Power Station Technical Specifications

prior to implementation of the improved Technical Specifications

(Amendment No. 95) and eliminates the potential for unnecessary power

reductions.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995.

Amendment No.: 100.

Facility Operating License No. NPF-62. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995. (60 FR

16190)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 2, 1995.

No significant hazard consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: July 22, 1993, as supplemented

February 4, August 23, September 16, October 6, and December 2, 1994,

and January 3, January 9, March 8, and April 10, 1995. [[Page 27350]]

Brief description of amendment: The amendment modified Facility

Operating License No. NPF-69 and the NMP-2 TSs to authorize an increase

in the maximum power level of NMP-2 from 3323 megawatts thermal

(MWt) to 3467 MWt. The amendment also approves changes to the

TSs to implement uprated power operation.

Date of issuance: April 28, 1995.

Effective date: As of the date of issuance to be implemented prior

to restart from refueling outage number 4.

Amendment No.: 66.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications and modifies Facility Operating License No.

NPF-69.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12360). The letters dated February 4, August 23, September 16, October

6, and December 2, 1994, and January 3, January 9, March 8, and April

10, 1995, provided clarifying information that did not change the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 28, 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: October 18, 1994, a supplemented

February 21, 1995.

Brief description of amendment: The amendment changes Surveillance

Requirement 4.6.1.2.a (Overall Integrated Containment Leakage Rate

Tests) by revising the surveillance interval for Type A tests from 40

plus or minus 10 months to approximately equal intervals during each

10-year inservice period. The amendment also removes a note that

expired upon completion of Cycle II refueling outage.

Date of issuance: May 3, 1995.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 187.

Facility Operating License No. DPR-65. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16191).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated may 3, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London turnpike, Norwich, CT 06360.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit no. 3, New London County, Connecticut

Date of application for amendment: December 23, 1994.

Brief description of amendment: The amendment changes the

acceptance criteria for the peak transient generator voltage from 4784

volts to 5000 volts during full load rejection tests of the diesel

generator (DG), and also deletes the 10-year surveillance requirement

to perform a 110% pressure test of the DG fuel oil system.

Date of issuance: May 1, 1995.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 110.

Facility Operating License No. NPF-49. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 15, 1995 (60

FR 8751).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 1, 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.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: September 28, 1994.

Brief description of amendment: The amendment revises Surveillance

Requirement 4.6.1.2.a of the Technical Specification to eliminate the

requirement to perform Type A tests on an interval of 40 plus or minus

10 months while reiterating the Appendix J requirement that the Type A

tests be performed three times, at approximately equal intervals,

during each 10 year service period. In addition, a footnote is added

which states that the third Type A test will be performed during the

sixth refueling outage. This reflects an exemption to Appendix J which

separates the third Type A test from the 10 year inservice inspection.

Date of issuance: May 8, 1995.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 111.

Facility Operating License No. NPF-49. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60384)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 8, 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 Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of application for amendment: August 19, 1994, as supplemented

March 15, 1995.

Brief description of amendment: The amendments add a new action

statement to Technical Specification 3.1.3.2.1., ``Position Indication

Systems--Operating''.

Date of issuance: May 3, 1995.

Effective date: May 3, 1995.

Amendment No.: 166 and 148.

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 12, 1994 (59 FR

51626) The March 15, 1995 supplement provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated may 3, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey 08079.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: March 19, 1993; superseded May

16, 1994; superseded February 10, 1995; supplemented February 17, 1995

(TS 93-04).

Brief description of amendment: The amendments clarify the Limiting

[[Page 27351]] Conditions for Operation applicable to the dual function

of the containment vacuum relief isolation lines by specifying the

actions that would be required should one or more of the vacuum relief

isolation lines by specifying the actions that would be required should

one or more of the vacuum relief lines be incapable of performing the

containment isolation function or incapable of performing the vacuum

relief function.

Date of issuance: April 28, 1995.

Effective date: April 28, 1995.

Amendment No.: 197 and 188.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: May 12, 1994 (58 FR

28060); renoticed June 22, 1994 (59 FR 32237), and March 29, 1995 (60

FR 16202).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 28, 1995.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendment: November 15, 1994; superseded

March 7, 1995 (TS 94-12).

Brief description of amendments: The amendments remove the

frequencies specified in the Technical Specifications for performing

audits and delete the requirement to perform the Radiological Emergency

Plan, Physical Security Plan, and Safeguard Contingency Plan reviews.

Date of issuance: May 10, 1995.

Effective date: May 10, 1995.

Amendment No.: 198 and 189.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: December 21, 1994 (59

FR 65823); renoticed March 29, 1995 (60 FR 16203)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 10, 1995.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: January 30, 1995.

Brief description of amendment: This amendment revises Technical

Specification (TS) 4.6.1.2.a, ``Containment Systems, Containment

Leakage, Surveillance Requirements (SR)'' and Bases 3/4.6,

``Containment Systems,'' to state that Type A tests for overall

integrated containment leakage rate testing shall be conducted in

accordance with the requirements specified in appendix J of 10 CFR part

50, as modified by NRC-approved exemptions. Additionally, TS SR

4.6.1.2.a.

Date of issuance: May 3, 1995.

Effective date: May 3, 1995.

Amendment No.: 198.

Facility Operating License No. NPF-3. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 15, 1995 (60 FR

14028).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 3, 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.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-

339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of application for amendments: July 8, 1993, as supplemented

by letters dated July 12, 1994, and March 7, 1995.

Brief description of amendments: The amendments revise the NA-1&2

Technical Specifications by deleting the requirements to periodically

review certain administrative and technical procedures.

Date of issuance: May 1, 1995.

Effective date: May 1, 1995.

Amendment Nos.: 190 and 171.

Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: August 4, 1993 (58 FR

41518).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 1, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-

339; North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of application for amendments: December 27, 1993, as

supplemented September 6, 1994, and March 7, 1995.

Brief description of amendments: The amendments revise the NA-1&2

Technical Specifications regarding the review responsibilities of the

Station Nuclear Safety and Operating Committee and the Management

Safety Review Committee.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995.

Amendment Nos.: 191 and 172.

Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: February 16, 1994 (59

FR 7700).

The September 6, 1994, and March 7, 1995 submittals provided

additional information only, and did not change the staff's initial

proposed determination of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of application for amendments: June 28, 1991.

Brief description of amendments: These amendments incorporate

operability and surveillance requirements for power-operated relief

valves to conform with Generic Letter 90-06.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995.

Amendment Nos.: 198 and 198.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 2, 1991 (56 FR

49929).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 2, 1995.

No significant hazards consideration comments received:

No. [[Page 27352]]

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Ch. 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 June 23, 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 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

[[Page 27353]] 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 lest

one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington, DC, by the above

date. Where petitions are filed during the last 10 days of the notice

period, it is requested that the petitioner promptly so inform the

Commission by a toll-free telephone call to Western Union 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).

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2

Date of application for amendments: April 24, 1995.

Brief description of amendments: the amendments change the

Technical Specifications by modifying the surveillance testing

periodicity requirements of the automatic actuation logic of engineered

safeguards equipment.

Date of issuance: May 5, 1995.

Effective date: May 5, 1995.

Amendment Nos.: 162 and 150.

Facility Operating Licenses Nos. DPR-39 and DPR-48. The amendments

revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: No.

The Commission's related evaluation of the amendments, finding of

emergency circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated May 5,

1995.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60690.

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

NRC Project Director: Robert A. Capra.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs

Nuclear Power Pant, Unit No. 1, Calvert County, Maryland

Date of application for amendment: April 28, 1995.

Brief description of amendment: The amendment revises the control

room emergency ventilation system TS 3.7.6.1, Limiting Condition For

Operation. The revision extends the one-time increase in the allowed

outage time for loss of emergency power only, from the 30 days

previously approved, to 45 days. This extension is necessary to allow

time to repair the Number 21 emergency diesel generator which failed

its operability tests subsequent to modifications which have been

recently completed.

Date of issuance: May 2, 1995.

Effective date: As of the date of issuance to be implemented upon

receipt.

Amendment No.: 205.

Facility Operating License No. DPR-53: Amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: No.

The Commission's related evaluation of the amendment, consultation

with the State, and final determination of no significant hazards

consideration are continued in a Safety Evaluation dated May 2, 1995.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

Potts and Trowbridge, 2300 N. Street, NW., Washington, DC 20037.

NRC Project Director: Ledyard B. Marsh.

Dated at Rockville, MD, this 17th day of May, 1995.

For the Nuclear Regulatory Commission,

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 95-12538 Filed 5-22-95; 8:45 am]

BILLING CODE 7590-01-M