Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from July 21, 1995, through August 4, 1995. The

last biweekly notice was published on Wednesday, August 2, 1995 (60 FR

39430).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By September 15, 1995, the licensee may file a request for a

hearing with respect to issuance of the amendment to the subject

facility operating license and any person whose interest may be

affected by this proceeding and who wishes to participate as a party in

the proceeding must file a written request for a hearing and a petition

for leave to intervene. Requests for a hearing and a petition for leave

to intervene shall be filed in accordance with the Commission's ``Rules

of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.

Interested persons should consult a current copy of 10 CFR 2.714 which

is available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if

[[Page 42598]]

proven, would entitle the petitioner to relief. A petitioner who fails

to file such a supplement which satisfies these requirements with

respect to at least one contention will not be permitted to participate

as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

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

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of amendments request: March 15, 1995, as supplemented on June

29, 1995.

Description of amendments request: The proposed amendments would

revise the Calvert Cliffs Nuclear Power Plant, Units Nos. 1 and 2,

Technical Specifications (TSs) Section 6, ``Administrative Controls,''

to be consistent with the guidance provided in NUREG-1432, ``Standard

Technical Specifications, Combustion Engineering Plants.'' The proposed

changes will relocate several requirements to other documents and

programs consistent with NUREG-1432 and other NRC guidance addressing

the administrative section of the TSs such as the ``Final Policy

Statement on Technical Specification Improvements for Nuclear Power

Reactors,'' published in the Federal Register on July 22, 1993 (58 FR

39132).

The Commission indicated that compliance with the Final Policy

Statement satisfies Section 182a of the Act. In particular, the

Commission indicated that certain items could be relocated from the TSs

to licensee-controlled documents, consistent with the standard

enunciated in Portland General Electric Co. (Trojan Nuclear Plant),

ALAB-531, 9 NRC 263, 273 (1979). In that case, the Atomic Safety and

Licensing Appeal Board indicated that ``technical specifications are to

be reserved for those matters as to which the imposition of rigid

conditions or limitations upon reactor operation is deemed necessary to

obviate the possibility of an abnormal situation or event giving rise

to an immediate threat to the public health and safety.'' The policy

statement encouraged licensees to adopt the applicable improved STSs

and provided some guidance for the conversion from the present plant-

specific TSs to the improved Standard TSs.

The proposed changes will provide significant human factors

improvement to the TSs by accomplishing the following: (1) relocating

existing requirements to licensee controlled documents consistent with

the policy statement; (2) eliminating requirements which duplicate

regulations; (3) relocating similar requirements within the same

section; (4) editorial changes; and (5) adding requirements consistent

with NUREG-1432.

In addition, the licensee proposes dual rolls for the Shift

Technical Advisor (STA) and the establishment of a TS Bases Control

Program. Allowing the STA to perform dual rolls is not permitted by the

current TSs, but the current NRC guidance allows the STA to perform a

dual roll. The proposed new TS Bases Control Program will define the

appropriate methods and reviews required to implement a TS Bases change

which is also consistent with the current NRC guidance. Two other

proposed changes, not specifically covered by the above groupings,

include a reduction in reporting requirements and utilizing a more

effective option for estimating doses.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

Relocating existing requirements to Baltimore Gas and Electric

Company (BGE)-controlled documents, eliminating requirements which

duplicate regulations, locating similar requirements within the same

sections and making necessary editorial corrections to incorporate

the proposed changes provide Technical Specifications which are

easier to use. Because existing requirements are relocated to

established BGE programs where changes to those programs are

controlled by regulatory requirements, there is no reduction in

commitment and adequate control is still maintained. Likewise, the

elimination of requirements which duplicate regulations enhances the

usability of the Technical Specifications without reducing

commitments. Locating similar requirements within the same sections

and making necessary editorial corrections to incorporate the

proposed changes neither add nor delete requirements, but merely

clarify and improve the readability and understanding of the

Technical Specifications. Since the requirements remain the same,

these changes only affect the method of presentation and would not

affect possible initiating events for accidents previously evaluated

or any system functional requirement. Therefore, the proposed

changes would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

[[Page 42599]]

Since the Shift Technical Adviser (STA) is not considered an

initiator to any previously evaluated accident nor considered in the

accident's response, the use of a dual role STA would not increase

the probability or consequences of any previously evaluated

accident.

The Technical Specification Bases Control Program provides

controls which ensure appropriate reviews of changes to the Bases.

Because NRC approval is still needed for changes to the Bases which

affect the Technical Specifications, the proposed Program would not

affect the probability or consequences of a previously evaluated

accident.

Eliminating the requirement for submitting two reports which

place unwarranted administrative burden on both Baltimore Gas and

Electric Company and the NRC has no affect on the probability or

consequences of an accident previously evaluated. Therefore, the

proposed changes would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Replacing the film badge with the electronic personal dosimeter

provides a more effective, efficient, state-of-the art option for

estimating dose and would not impact accidents previously evaluated.

Therefore, the proposed change would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

As discussed previously, relocating existing requirements to

BGE-controlled documents, eliminating requirements which duplicate

regulations, locating similar requirements within the same sections

and making necessary editorial corrections to incorporate the

proposed changes will not affect any plant system or structure, nor

will it affect any system functional or operability requirements.

Consequently, no new failure modes are introduced as a result of the

proposed changes. Therefore, these types of changes would not create

the possibility of a new or different type of accident from any

accident previously evaluated.

Because the STA does not perform equipment design or equipment

manipulation, the use of a dual role STA would not create the

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

previously evaluated. Since the Technical Specification Bases

Control Program represents an administrative function performed

under existing regulatory controls, it too would not create the

possibility of a new or different type of accident from any

previously evaluated.

The addition of new programs which incorporate existing

Technical Specification requirements and commitments will have no

effect on the design or operation of the plant and would not create

the possibility of a new or different type of accident from any

previously evaluated.

A reporting function such as report submittals would not change

the configuration or operation of the plant. Consequently, the

elimination of the requirement to submit the Startup Report and the

Special Report dealing with iodine activity levels, would not create

the possibility of a new or different type of accident from any

accident previously evaluated.

Since the operation or configuration of the plant is not changed

by the type of personal dosimeter, this change would not create the

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

previously evaluated.

Therefore, the proposed changes would not create the possibility

of a new or different type of accident from any accident previously

evaluated.

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

safety.

Relocating existing requirements to BGE-controlled documents,

eliminating requirements which duplicate regulations, locating

similar requirements within the same sections and making necessary

editorial corrections to incorporate the proposed changes would not

affect the Updated Final Safety Analysis Report design bases,

accident analysis assumptions or any margin of safety described in

the Technical Specification Bases. In addition, these proposed

changes do not affect effluent release limits, monitoring equipment

or practices. Therefore, these proposed changes would not involve a

significant reduction in a margin of safety.

The use of an STA should provide an additional margin of safety

in the accident response function of licensed operators beyond that

considered in the accident analysis. Since the STA is required to

have the same training and educational qualifications in either the

individual or dual role, the use of a dual role STA should have

minimal impact. Consequently, the proposed change would not involve

a significant reduction in a margin of safety. The Technical

Specification Bases Control Program is an administrative change

controlling how Technical Specification basis information is

reviewed and incorporated. Therefore, this change would not involve

a significant reduction in a margin of safety.

The addition of new programs which incorporate existing

Technical Specification requirements and commitments will have no

effect on the design or operation of the plant and would not result

in a significant reduction in the margin of safety.

Activities described in the Startup Report will continue to be

performed and corrective action taken when required. Similarly,

iodine activity levels will continue to be monitored and actions

taken, including the issuance of a Licensee Event Report when

conditions warrant. Considering the above, elimination of the two

reporting requirements would have no impact on the margin of safety.

Plant operating parameters are not affected by the type of

personnel monitoring device used and as a consequence, would not

impact a margin of safety. Since the replacement dosimeter provides

a more effective mechanism for estimating dose, there is no

degradation in personal safety levels. Consequently, the proposed

change would not involve a significant reduction in a margin of

safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

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

NRC Project Director: Ledyard B. Marsh

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendment requests: September 17, 1993, as

supplemented July 28, 1995

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

Diagnostic Evaluation Team inspection performed by the NRC staff at the

Dresden Nuclear Power Station in 1987, Commonwealth Edison Company

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

Power Station and sister site Quad Cities Nuclear Power Station needed

attention focused on the existing custom Technical Specifications (TS)

used.

The licensee made the decision to initiate a Technical

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

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

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

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

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

improvements such as clarifying requirements, changing TS to make them

more understandable and to eliminate interpretation, and deleting

requirements that are no longer considered current with industry

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

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

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

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

such as equipment operability requirements during shutdown conditions,

(2) clarifying requirements such as limiting conditions for operation

and action

[[Page 42600]]

statements utilizing STS terminology, (3) deleting superseded

requirements and modifications to the TS based on the licensee's

responses to Generic Letters (GL), and (4) relocating specific items to

more appropriate TS locations.

The September 17, 1993, and July 28, 1995, applications proposed to

upgrade only Section 3/4.5 (Emergency Core Cooling Systems) of the

Dresden and Quad Cities TS.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

In general, the proposed amendment represents the conversion of

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

requirements which are based on the current safety analysis.

Implementation of these changes will provide increased reliability

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

provide continued assurance that specified parameters remain within

their acceptance limits, and as such, will not significantly

increase the probability or consequences of a previously evaluated

accident.

Some of the proposed changes represent minor curtailments of the

current requirements which are based on generic guidance or

previously approved provisions for other stations. The proposed

amendment for Dresden and Quad Cities Station's Technical

Specification Section 3/4.5 are based on STS guidelines or later

operating BWR plants' NRC accepted changes. Any deviations from STS

requirements do not significantly increase the probability or

consequences of any previously evaluated accidents for Dresden or

Quad Cities Stations. The proposed amendment is consistent with the

current safety analyses and has been previously determined to

represent sufficient requirements for the assurance and reliability

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

continued assurance that specified parameters remain within their

acceptance limits. As such, these changes will not significantly

increase the probability or consequences of a previously evaluated

accident.

The associated systems that make up the Emergency Core Cooling

Systems are not assumed in any safety analysis to initiate any

accident sequence for Dresden or Quad Cities Stations; therefore,

the probability of any accident previously evaluated is not

increased by the proposed amendment. In addition, the proposed

surveillance requirements for the proposed amendments to these

systems are generally more prescriptive than the current

requirements specified within the Technical Specifications. The

additional surveillance requirements improve the reliability and

availability of all affected systems and therefore, reduce the

consequences of any accident previously evaluated as the probability

of the systems outlined within Section 3/4.5 of the proposed

Technical Specifications performing their intended function is

increased by the additional surveillances.

Create the possibility of a new or different kind of accident

from any previously evaluated because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, the addition of

requirements which are based on the current safety analysis, and

some minor curtailments of the current requirements which are based

on generic guidance or previously approved provisions for other

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

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

operation of the station; however, these provide additional

restrictions which are in accordance with the current safety

analysis, or are to provide for additional testing or surveillances

which will not introduce new failure mechanisms beyond those already

considered in the current safety analyses.

The proposed amendment for Dresden and Quad Cities Station's

Technical Specification Section 3/4.5 is based on STS guidelines or

later operating BWR plants' NRC accepted changes. The proposed

amendment has been reviewed for acceptability at the Dresden and

Quad Cities Nuclear Power Stations considering similarity of system

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

deviations from STS requirements do not create the possibility of a

new or different kind of accident previously evaluated for Dresden

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

the proposed changes. Surveillance requirements are changed to

reflect improvements in technique, frequency of performance or

operating experience at later plants. Proposed changes to action

statements in many places add requirements that are not in the

present technical specifications. The proposed changes maintain at

least the present level of operability. Therefore, the proposed

changes do not create the possibility of a new or different kind of

accident from any previously evaluated.

The associated systems that make up the Emergency Core Cooling

Systems are not assumed in any safety analysis to initiate any

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

the proposed surveillance requirements for affected systems

associated with the Emergency Core Cooling Systems are generally

more prescriptive than the current requirements specified within the

Technical Specifications; therefore, the proposed changes do not

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

any previously evaluated.

Involve a significant reduction in the margin of safety because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, the addition of

requirements which are based on the current safety analysis, and

some minor curtailments of the current requirements which are based

on generic guidance or previously approved provisions for other

stations. Some of the latter individual items may introduce minor

reductions in the margin of safety when compared to the current

requirements. However, other individual changes are the adoption of

new requirements which will provide significant enhancement of the

reliability of the equipment assumed to operate in the safety

analysis, or provide enhanced assurance that specified parameters

remain with their acceptance limits. These enhancements compensate

for the individual minor reductions, such that taken together, the

proposed changes will not significantly reduce the margin of safety.

The proposed amendment to Technical Specification Section 3/4.5

implements present requirements, or the intent of present

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

Any deviations from STS requirements do not significantly reduce the

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

changes are intended to improve readability, usability, and the

understanding of technical specification requirements while

maintaining acceptable levels of safe operation. The proposed

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

Dresden or Quad Cities based on system design, safety analysis

requirements and operational performance. Since the proposed changes

are based on NRC accepted provisions at other operating plants that

are applicable at Dresden or Quad Cities and maintain necessary

levels of system or component reliability, the proposed changes do

not involve a significant reduction in the margin of safety.

The proposed amendment for Dresden and Quad Cities Stations will

not reduce the availability of systems associated with the Emergency

Core Cooling Systems when required to mitigate accident conditions;

therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,

Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

[[Page 42601]]

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

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

Date of amendment request: June 17, 1993, as supplemented July 5,

1995

Description of amendment request: The initial proposed amendment

request dated June 17, 1993, was previously noticed in the Federal

Register on July 21, 1993 (58 FR 39048). The proposed amendment would

revise Technical Specification 5.3.1, ``Fuel Assemblies'' to provide

flexibility in the repair of fuel assemblies containing damaged and

leaking fuel rods by reconstituting the assemblies in accordance with

the guidance in Generic Letter (GL) 90-02, Supplement 1, ``Alternative

Requirements For Fuel Assemblies In The Design Features Section Of

Technical Specifications,'' issued on July 31, 1992. The application is

also generally consistent with the format and content of the improved

Standard Technical Specifications for Westinghouse plants provided in

NUREG-1431.

Additional information was submitted on July 5, 1995, that added TS

changes to increase the fuel enrichment limit from 4.0 to 5.0 weight

percent U-235 that were not previously included the initial June 17,

1993, amendment application. This additional information is being

noticed to provide for public comment and opportunity for hearing.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee provided its analysis of the issue of no significant hazards

consideration (58 FR 39048). The NRC staff's analysis of the July 5,

1995, supplement against the standards of 10 CFR 50.92(c) is presented

below.

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

the probability or consequences of an accident previously evaluated.

There is no increase in the probability or consequences of an

accident in the new fuel vault since the only accident that would be

affected by this change would be a criticality accident and it has been

shown that the worst-case keff under optimum moderation conditions

continues to be less than or equal to 0.98.

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

the Spent Fuel Storage Pool since the mass of an assembly will not be

significantly affected by the increase in fuel enrichment. The

likelihood of other accidents, previously evaluated and described in

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

affected by the proposed changes. Since the increase in fuel enrichment

will allow for extended fuel cycles, it could be postulated that there

may be a decrease in fuel movement and the probability of an accident

may likewise be decreased. There is also no increase in the

consequences of a fuel drop accident in the Spent Fuel Pool since the

fission product inventory of individual fuel assemblies will not change

significantly as a result of increased initial enrichment. In addition,

no change to safety-related systems is being made.

Therefore, the consequences of a fuel rupture accident remain

unchanged. In addition, it has been shown that keff is less than

or equal to 0.95, under all conditions. Therefore, the consequences of

a criticality accident in the Spent Fuel Pool remain unchanged as well.

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

or different kind of accident from any accident previously

evaluated.

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

different kind of accident since fuel handling accidents (fuel drop and

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

handling accidents are already discussed in the FSAR for fuel with

enrichments up to 4.0 weight % and additional analyses have been

performed for fuel with enrichment up to 5.00 weight %.

3.

The proposed changes do not involve a significant reduction in the

margin of safety.

The proposed change does not involve a significant reduction in the

margin of safety since, in all cases, a spent fuel pool keff less

than or equal to 0.95 is being maintained. Criticality analyses have

also been performed that show that the new fuel storage vault will

remain subcritical under a variety of moderation conditions, from fully

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

will remain sufficiently subcritical during any fuel misplacement

accident.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to

determine that the supplemental amendment submittal involves no

significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: July 26, 1995

Description of amendment request: The proposed amendments would

provide a one-time extension of the allowable outage time from 72 hours

to 7 days. This extension is necessary to implement a modification to

the degraded grid protection system and the external grid trouble

protection system.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

Duke Power Company (Duke) has made the determination that this

amendment request involves a No Significant Hazards Consideration

by applying the standards established by NRC regulations in 10 CFR

50.92. This ensures that operation of the facility in accordance

with the proposed amendment would not:(1) Involve a significant

increase in the probability or consequences of an accident

previously evaluated:

Each accident analysis addressed within the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to the change

proposed within this amendment request. The design basis of the

auxiliary electrical systems is to supply the required engineered

safeguards (ES) loads of one unit and the safe shutdown loads of the

other two units. The systems are arranged so that no single failure

will jeopardize plant safety.

The probability of any Design Basis Accident (DBA) is not

significantly increased by this change. In addition, the

consequences of the accidents are within the bounds of the FSAR

analyses. The reliability of the emergency power system is not

significantly affected by a one time extension of allowable outage

time for the overhead power path. The underground power path is

adequate to assure operability of the Oconee ES loads. Finally, the

enhancement of the Degraded [Grid] Protection System will eliminate

a concern which was expressed by the EDSFI audit team.

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

accident from any kind of accident previously evaluated:

Inoperability of the yellow bus is functionally equivalent to

inoperability of the Keowee Main Step-up Transformer in that it

renders the overhead emergency power path inoperable. The Keowee

Main Step-up Transformer is allowed to be inoperable for a period

not to exceed 28 days. This Technical Specification requirement for

the

[[Page 42602]]

Keowee Main Step-up Transformer has been reviewed and approved by the

NRC. Therefore, operation of ONS [Oconee Nuclear Station] in

accordance with this Technical Specification amendment will not

create any failure modes not bounded by previously evaluated

accidents. Consequently, this change will not create the possibility

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

previously evaluated.

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

The design basis of auxiliary electrical systems is to supply

the required ES loads of one Unit and safe shutdown loads of the

other two units. The underground power path is adequate to ensure

operability of the ES loads during the outage of the yellow bus. The

reliability of the emergency power system is not significantly

affected by a one time extension of allowable outage time for the

overhead power path. Therefore, there will be no significant

reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: July 10, 1995

Description of amendment request: The proposed amendment would

modify the technical specifications to minimize the potential for boron

deletion of the reactor coolant system (RCS) during startup of an

isolated loop. The changes would permit RCS loop isolation only during

Modes 5 and 6. RCS loop isolation valves would be required open with

power removed from each isolation valve operator during Modes 1, 2, 3,

and 4. Primary grade water would be isolated from the RCS during Modes

4, 5, and 6, except during planned boron dilution or makeup activities.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed amendment would modify the method used to prevent

an inadvertent boron dilution event during hot shutdown, cold

shutdown and during refueling. An uncontrolled boron dilution

transient cannot occur during this mode of operation. Inadvertent

boron dilution is prevented by administrative controls which isolate

the primary grade water system isolation valves from the Chemical

and Volume Control System, except during planned boron dilution or

makeup activities. Thus unborated water can not be injected into the

reactor coolant system, making an unplanned boron dilution at these

conditions highly improbable, since the source of unborated water to

the charging pumps is isolated. This precludes the primary means for

an inadvertent boron dilution event in this mode of operation.

The primary grade water system isolation valves may be opened

when directed by the control room during this mode of operation only

for a planned boron dilution or makeup activity. The primary grade

water system isolation valves will be verified to be locked, sealed

or otherwise secured in the closed position after the planned boron

dilution or makeup activity is completed. During planned boron

dilution events, operator attention will be focused on the boron

dilution process and any inappropriate blender operation will be

readily identified.

The operator has prompt and definite indication of any boron

dilution from the audible count rate instrumentation supplied by the

source range nuclear instrumentation. High count rate is alarmed in

the reactor containment and the control room. In addition a high

source range flux level is alarmed in the control room. The count

rate increase is proportional to the subcritical multiplication

factor.

The proposed amendment would also modify the method used to

prevent an adverse reactor transient during startup of an isolated

reactor coolant loop. Procedures require that the isolated loop

water boron concentration be verified prior to opening loop

isolation valves. Procedures also require an isolated loop to be

drained and refilled from water supplied from the Refueling Water

Storage Tank (RWST) or Reactor Coolant System (RCS) prior to opening

either the hot or cold leg isolation valves. Using water from the

RWST or RCS ensures 1) that the boron concentration of the isolated

loop is sufficient to prevent a dilution of the active reactor

coolant loops and reducing the shutdown margin to below those values

used in safety analyses when the isolated loop is returned to

service, and 2) that no single failure could cause an isolated loop

to be filled with unborated water.

Thus procedures and interlocks prevent inadvertent opening of

loop isolation valves and require that the startup of an isolated

loop be performed in a controlled manner that virtually eliminates

any sudden positive reactivity addition from boron dilution. Thus

the core cannot be adversely affected by the startup of an isolated

loop and fuel design limits are not exceeded. Therefore, the

proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

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

different kind of accident. No new systems, structures or components

are being proposed. Acceptable alternative administrative controls

are being proposed to address inadvertent boron dilution and the

startup of inactive reactor coolant loops.

The primary source of unborated water will be isolated from

injecting by the charging pumps into the reactor coolant system

during hot shutdown, cold shutdown, and refueling, except for

planned boron dilution events and makeup activities. The proposed

administrative controls prevent the possible accident previously

evaluated, i.e., an inadvertent boron dilution event.

A currently installed interlock to recirculate reactor coolant

in an isolated loop is proposed to be deleted. In its place, each

reactor coolant isolated loop will be drained and refilled with

water supplied from the RWST just before the loop is returned to

service. This administrative control will prevent any inadvertent

reactivity transient when returning the loop to service. Thus, the

proposed administrative controls will prevent the type of accident

previously evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes will continue to ensure that adequate

protection is provided against an inadvertent boron dilution and the

adverse effects from the startup of an isolated reactor coolant

loop. General Design Criteria 10 requirements will not be exceeded

with respect to demonstrating specified acceptable fuel design

limits. The required indications and functions are still maintained

in accordance with current technical specification requirements and

the shutdown margin is unaffected. Therefore, the proposed change

will not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: B. Library, 663 Franklin

Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: John F. Stolz

[[Page 42603]]

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley

Power Station, Unit 1, Shippingport, Pennsylvania

Date of amendment request: July 11, 1995

Description of amendment request: The proposed amendment would

revise the required area of the Reactor Coolant System (RCS)

overpressure protection system vent from 3.14 square inches to 2.07

square inches. This vent is provided to relieve a potential RCS

overpressure condition if the power-operated relief valves (PORVs) are

not operable. The proposed vent area is equal to the relief area of a

PORV. A single PORV is capable of providing sufficient relief capacity

to mitigate potential low temperature overpressurization events.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change is considered to be editorial since it

replaces the 3.14 square inch vent size stated in overpressure

protection system (OPPS) Specifications 3.4.9.3, 3.1.2.1.b, and

3.1.2.3 and Bases 3/4.1.2 and 3/4.4.9 with a 2.07 square inch vent

size. This ensures the vent size stated in the technical

specifications is consistent with the actual size of an installed

PORV. These changes maintain consistency with the analyses

assumptions and the operation of the OPPS in accordance with

applicable analyses and the UFSAR [Updated Final Safety Analyses

Report]. Therefore, we have concluded that these changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated in the UFSAR.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes do not involve any physical changes to the

OPPS or their setpoints. These changes do not change any function

previously provided by the OPPS. These changes do not affect any

failure modes defined for any plant system or component important to

safety nor has any new limiting single failure been identified as a

result of these changes. Therefore, these changes will not create

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

accident previously evaluated in the UFSAR.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes will not affect the operation of or the

reliability of the OPPS. These changes do not affect the manner in

which the plant is operated or involve a change to equipment or

features that affect the operational characteristics of the plant.

Therefore, operation of the plant in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: July 20, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.8.1.1 to incorporate guidance

provided in NRC Generic Letter (GL) 84-15, ``Proposed Staff Actions to

Improve and Maintain Diesel Generator Reliability,'' and GL 93-05,

``Line-Item Technical Specification Improvements To Reduce Surveillance

Requirements For Testing During Power Operation,'' which includes (1)

revised requirements for testing the operable emergency diesel

generators (EDGs) for various combinations of inoperable offsite

circuits and EDGs and (2) revised surveillance requirements for the

EDGs. The revised surveillance requirements include specifying

generator voltage, frequency limits, and diesel starting time. In

addition, several editorial changes would be made to TS 3/4.8.1.1 which

would be consistent with the guidance provided in the NRC's Improved

Standard Technical Specifications (NUREG-1431).

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The probability of occurrence of a previously evaluated accident

is not increased because the allowable outage times for the offsite

circuits and diesel generators remain unchanged. The consequences of

an accident previously evaluated is not increased because reducing

the diesel generator test frequency and permitting additional test

evolutions are intended to minimize diesel wear and mechanical

stress. By eliminating excessive testing, which can lead to

premature diesel failures and minimizing diesel wear and mechanical

stress, the diesel generator reliability is increased. The

consequences of an accident previously evaluated is also not

increased because the addition of the parameters for generator

voltage, frequency, and diesel starting time to the surveillance

requirement will provide additional assurance that the diesel

generators are performing as assumed in the safety analysis. This

proposed change does not affect the availability or reliability of

the offsite circuits.

Therefore, this change will not increase the probability or

consequences of an accident previously evaluated due to the

continued availability and reliability of the A.C. electrical power

sources.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes do not alter the method of operating the

plant. The changes do not introduce any new failure modes and are

intended to increase the diesel generator reliability and provide

additional assurance that the diesels are performing as assumed in

the safety analysis. The revision to the various action statements

and surveillance requirements provide assurance that the diesel

generators will be able to power their respective safety systems if

required. The proposed changes do not impact the performance of any

safety system.

Therefore, this proposed change does not create the possibility

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

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The margin of safety is not reduced because the A.C. electrical

power sources will continue to provide sufficient capacity,

capability, redundancy, and reliability to ensure availability of

necessary power to engineered safety feature (ESF) systems. The ESF

systems will continue to function, as assumed in the safety

analyses, to ensure that the fuel, reactor coolant system and

containment design limits are not exceeded. The elimination of

excessive testing on the diesel generators are permitting additional

test evolutions, which result in less diesel wear and mechanical

stress, are intended to increase diesel reliability. The increased

reliability of the diesels adds to the ability of the A.C.

electrical power source to provide power to ESF systems. The

proposed additions to the surveillance requirements will provide

additional assurance of the ability of the A.C. electrical power

sources to provide power to ESF systems.

Therefore, this proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

[[Page 42604]]

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: John F. Stolz

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

PowerStation, Unit 2, Shippingport, Pennsylvania

Date of amendment request: July 24, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.4.11, ``Relief Valves,'' and

associated Bases to make Unit 2 TS 3/4.4.11 consistent with Unit 1 TS

3/4.4.11, which was revised by Unit 1 License Amendment No. 187 issued

on May 15, 1995. The proposed amendment would also generally reflect

the guidance provided in NRC Generic Letter 90-06 and in the NRC's

Improved Standard Technical Specifications (NUREG-1431).

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Implementation of these changes will increase the availability

of the power-operated relief valves (PORVs) and their associated

block valves. The increased availability is obtained through

maintaining power to the block valves which are closed to control

PORV seat leakage. Maintaining power to the block valve provides the

flexibility of reopening the valves to control reactor coolant

system pressure. The proposed change modifies Specification 3.4.11

actions, a surveillance requirement, and Bases to generally reflect

the requirements of Generic Letter (GL) 90-06, and the guidance

provided in NUREG-1431, ``Improved Standard Technical

Specifications'' (ISTS) and is consistent with the changes the NRC

approved for Unit No. 1. A revised stress analysis has been

completed that takes credit for the speed at which the block valve

opens when manually reducing reactor coolant system pressure. The

block valve relatively slow opening speed reduces the peak pressure

surge and results in acceptable downstream piping stress values. The

PORV downstream piping has been evaluated assuming manual vent path

operation with cold loop seal slug flow and it has been determined

that the piping supports can accept these design transient loads.

The proposed change to the action statement to close the block valve

to isolate a PORV and maintain power to the block valve does not

significantly increase the probability of a small break loss of

coolant accident. No PORV function has been deleted and the PORV and

block valve continue to be capable of being manually closed at any

time. As a result of the change to action ``a,'' an exception to the

stroking requirements is no longer required, therefore, reference to

action ``a'' in Surveillance Requirement 4.4.11.2 has been deleted.

Closing the block valve for a PORV that is not capable of being

manually cycled and removing power to the block valve assures that

the valve will not be inadvertently opened when the condition of the

PORV is uncertain.

The changes remain consistent with the analysis assumptions

regarding the operation of the PORVs and block valves and provides

increased assurance of their availability in mitigating the

consequences of a steam generator tube rupture (SGTR) accident. The

requirements of GL 90-06 are substantially addressed in the ISTS

which have been incorporated here except for specific design

differences. Minor editorial changes involving capitalization have

been incorporated to maintain the format and content and do not

affect any of the requirements, the accident analyses, or the

operation of the plant. Therefore, we have concluded that these

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated in the UFSAR

[Updated Final Safety Analysis Report].

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes to the action statements for the PORVs and

the associated block valves will improve the availability of these

valves for normal operation and for mitigation of a SGTR accident.

The proposed changes do not involve any physical changes to the

PORVs or their setpoints. These changes do not delete any design

basis accident function previously provided by the PORV vent path

nor has the probability of inadvertent opening been increased.

Accordingly, no new limiting single failure has been identified as a

result of these changes. Therefore, these changes will not create

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

accident previously evaluated in the UFSAR.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes have been incorporated to provide the

capability to manually stroke the vent path using the block valve to

control the pressure surge as a PORV opens. The resultant downstream

piping forces were found acceptable, therefore, power can be

maintained to the block valve when the block valve has been closed

to isolate a PORV because of excessive seat leakage. This will allow

operation of the PORVs in a manner similar to the guidance provided

in GL 90-06 to improve PORV availability. These changes will improve

the operator use of an isolated PORV since it is now analyzed to be

manually cycled with the block valve closed and power maintained so

the operator can use the PORV if required to mitigate the effects of

a SGTR accident. This is consistent with the intent of the ISTS and

does not affect the UFSAR, therefore, operation of the plant in

accordance with the proposed amendment would not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,

Unit No. 2, Pope County, Arkansas

Date of amendment request: April 4, 1995

Description of amendment request: The proposed amendment revises

the minimum water level that is required to be maintained over

irradiated fuel assemblies during latching and unlatching of control

element assemblies.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The fuel handling accident analysis assumes that a fuel assembly

is dropped during fuel handling. During the latching and unlatching

of the CEAs, the upper guide structure is in place and the CEDM

extension shaft assemblies are disconnected from their CEA for

subsequent removal with the vessel upper guide structure. The

dropping of a CEA from the maximum height of six inches will not

damage that particular fuel assembly or any surrounding fuel

assemblies since this movement is confined to within the upper guide

structure and the guide tubes of the associated fuel assembly during

this activity. This less than six inches of movement does not have

the potential to result in a fuel handling accident; therefore, an

increase in the probability of this accident does not occur. The

requirement to have at least 23 feet of water over the top of the

irradiated fuel assemblies during fuel and CEA movement ensures

that, should a fuel handling accident occur, the resulting offsite

dose consequences are mitigated. The six inch movement of the CEA

during CEA decoupling does not constitute fuel or CEA

[[Page 42605]]

movement which would result in a fuel handling accident. As such,

Technical Specifications are unchanged with respect to the

mitigating requirements for a fuel handling accident.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The proposed change does not change the design, configuration,

or method of operation of the plant; therefore, it does not create

the possibility of a new or different kind of accident. Because no

new equipment is being introduced, and no equipment is being

operated in a manner inconsistent with its design, the possibility

of equipment malfunction is not increased. The proposed change adds

an exception to the applicability section and is bounded by the

existing fuel handling accident analysis.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

There is no reduction in margin of safety in that 23 feet of

water is still maintained over the irradiated fuel assemblies

anytime there is a potential for a fuel handling accident. Adding

the exception of the latching and unlatching of the CEAs to the

applicability section does not involve a change in the accident

analysis for fuel handling which remains bounding.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: July 21, 1995

Description of amendment request: The proposed change requests that

the current expiration date for license NPF-29 be changed to reflect

the issuance date of the new license granted Grand Gulf on November 1,

1984. The change consists of extending the expiration date to 40 years

from the date of issuance of license NPF-29 (November 1, 1984 to

November 1, 2024).

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

a. No significant increase in the probability or consequences of

an accident previously evaluated results from this change.

The proposed change does not affect the design or operation of

any plant system. The effect of 40 years of full power operations

has previously been evaluated and documented in the Updated Final

Safety Analysis Report (UFSAR). The design life of structures,

systems and components is controlled by existing plant problems

[sic., programs] and processes that are not affected by this change.

The proposed change will simply allow Grand Gulf to achieve its

original planned 40 years of service. Equipment associated with

initiating event frequencies or accident mitigation must continue to

meet all applicable maintenance and operability requirements

regardless of license duration (It is also interesting to note that

the license duration limitation of 40 years, as contained in 10 CFR

50.51 is not a limitation resulting from concerns over plant aging

effects. ``In fact, the limit was a compromise between the efforts

of the Justice Department and electric cooperatives, who championed

a 20-year limit on the basis of antitrust concerns, and the view of

the utility industries that a longer period was necessary to ensure

full amortization of a nuclear power plant.'' (56 FR 64961, December

13, 1991)). Therefore, the probability or consequences of previously

analyzed accidents are not significantly increased.

b. The change would not create the possibility of a new or

different kind of accident from any previously analyzed.

The proposed change will not add any plant equipment or

introduce any new modes of plant operation. The change will only

amend the operating license to allow 40 years of full power

operations. The proposed change does not affect the current

maintenance or surveillance practices, which are designed to

maintain and monitor the current service life of plant structures,

systems and components in accordance with regulatory requirements.

Therefore, the proposed change does not create the possibility of

new equipment failure modes or a new or different kind of accident

from any accident previously evaluated.

c. The change would not involve a significant reduction in a

margin of safety.

The proposed change does not involve a significant reduction in

a margin of safety since it only provides for 40 years of full power

operations for which the plant is designed. Current Technical

Specification surveillance requirements (e.g. associated with 10 CFR

50 Appendix H) and other regulatory requirements remain in place and

will ensure continued compliance with applicable safety margins.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, Mississippi 39120

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502

NRC Project Director: William D. Beckner

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania

Date of amendment request: June 20, 1995

Description of amendment request: The proposed Technical

Specifications (TS) changes would remove the surveillance interval text

for the 10 CFR Part 50, Appendix J, Type A test (Integrated Leak Rate

Test or ILRT), and Drywell-to-Suppression Chamber (bypass) leakage test

specified in TS Surveillance Requirements (SR) 4.6.1.2.a, 4.6.1.2.b,

and 4.6.2.1.e.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The proposed TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The primary containment and the suppression chamber are not

considered to be accident initiators, they are accident mitigators.

There are no physical or operational changes to the containment or

suppression structure, system or components being made as a result

of the proposed changes. These changes will not impose different

requirements and adequate control of information will be maintained.

These TS changes will not alter assumptions made in the safety

analysis and licensing basis. Therefore, the proposed TS changes to

eliminate the details of the test intervals will not increase the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes remove the specific surveillance test

interval text from TS and address the interval by direct reference

to the applicable regulation. The proposed TS changes do not make

any physical or operational changes to existing plant systems or

components. Furthermore, the primary containment and suppression

chamber act as

[[Page 42606]]

accident mitigators not initiators. Therefore, the possibility of a new

or different kind of accident than from any accident previously

evaluated is not introduced.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

LGS [Limerick Generating Station] TS Bases 3/4 6.1.2 state that

surveillance testing is consistent with 10 CFR 50, Appendix J and

does not specify a SR test interval. TS Bases 3/4 6.2, describing

the bypass test does not specify a SR test interval. However, the

NRC Safety Evaluation related to amendment Nos. 68 (Unit 1) and 31

(Unit 2) concluded that it is acceptable for the drywell-to-

suppression chamber test frequency to coincide with the 10 CFR 50,

Appendix J, Type A test, since individual vacuum breaker leakage

tests are an acceptable alternative to an integrated suppression

pool bypass test during outages for which a Type A containment

integrated leak rate test is not conducted. The alternative bypass

test requirement, TS SR 4.6.2.1.f, is not affected by these changes.

The Type A test, and bypass SR test intervals are adequately

presented in the test implementing procedures, and TS will directly

reference 10 CFR 50, Appendix J, for the appropriate test interval.

Therefore, the proposed TS changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Power Authority of The State of New York, Docket No. 50-286, Indian

PointNuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: July 21, 1995

Description of amendment request: The proposed amendment would

change Technical Specifications Section 6.0 (Administrative Controls)

to replace the title-specific list of members on the Plant Operating

Review Committee (PORC) with a more general statement of membership

requirements. The scope of disciplines represented on the PORC would

also be expanded to include nuclear licensing and quality assurance.

The proposed amendment would also change the title ``Resident Manager''

to ``Site Executive Officer.'' This title change would not affect the

reporting relationship, authority, or responsibility of the position.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

Operation of the Indian Point 3 Nuclear Power Plant in

accordance with the proposed amendment would not involve a

significant hazards consideration as defined in 10 CFR 50.92, since

it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are administrative in nature and do not

involve plant equipment or operating parameters. There is no change

to any accident analysis assumptions or other conditions which could

affect previously evaluated accidents. The proposed changes will not

decrease the organization's ability to respond to a design basis

accident.

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

from those previously evaluated.

Since the proposed changes are administrative in nature and do

not involve hardware design, modifications or operation, the

possibility of new or different accidents is not created.

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

The proposed title change for the Resident Manager is an

administrative change and does not affect the responsibilities,

authority, or reporting relationships for this management position.

Replacing the title specific list of PORC members with a statement

of membership requirements for the committee does not reduce the

effectiveness of the committee to advise the Resident Manager (Site

Executive Officer) on matters regarding nuclear safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Ledyard B. Marsh

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 30, 1995

Description of amendment request: The proposed change to the

Technical Specifications (TS) would change TS Table 3.3.1-2, ``Reactor

Protection System Response Times'', TS Table 3.3.2-3, ``Isolation

System Instrumentation Response Time'', TS Table 3.3.3-3, ``Emergency

Core Cooling System Response Times'', and associated Bases. The

proposed changes to the above-referenced TS Tables would eliminate the

requirement to perform response time testing for certain classes of

equipment.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

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

consequences of an accident previously evaluated.

The purpose of the proposed Technical Specification change is to

eliminate response time testing requirements for selected

instrumentation in the Reactor Protection System, Isolation System,

and Emergency Core Cooling System. However, because of the continued

application of other existing Technical Specification requirements

such as channel calibrations, channel checks, channel functional

tests, and logic system functional tests, the response time of these

systems will be maintained within the acceptance limits assumed in

plant safety analyses and required for successful mitigation of an

initiating event. The proposed Technical Specification changes do

not affect the capability of the associated systems to perform their

intended function within their required response time.

The BWR Owners' Group has completed an evaluation (NEDO-32291,

``System Analyses for the Elimination of Selected Response Time

Testing Requirements'') which demonstrates that response time

testing is redundant to the other Technical Specification

requirements listed in the preceding paragraph. These other tests

are sufficient to identify failure modes or degradation in

instruments response time and ensure operation of the associated

systems within acceptance limits. There are no known failure modes

that can be detected by response time testing that cannot be

detected by the other Technical Specification tests. Hope Creek

Generating Station is specifically bounded by the assumptions and

justifications in General Electric Company Licensing Topical Report,

NEDO-32291, ``System Analyses for Elimination of Selected Response

Time Testing Requirements.''

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

accident from any accident previously evaluated.

As discussed above, the proposed Technical Specification changes

do not affect the capability of the associated systems to perform

their intended function within the acceptance limits assumed in

plant safety analyses and required for successful mitigation of an

initiating event. The proposed elimination of response time testing

would not result in any new

[[Page 42607]]

equipment, operating modes, or plant configurations.

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

safety.

The current Technical Specification response times are based on

the maximum allowable values assumed in the plant safety analyses.

These analyses conservatively establish the margin of safety. As

described above, the proposed Technical Specification changes do not

affect the capability of the associated systems to perform their

intended functions within the allowed response time used as the

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

initiating event will remain in compliance within the assumptions of

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

affected.

Although not explicitly evaluated, the proposed Technical

Specification changes will provide an improvement to plant safety

and operation by:

a) Reducing the time safety systems are unavailable

b) Reducing safety system actuations

c) Reducing shutdown risk

d) Limiting radiation exposure to plant personnel

e) Eliminating the diversion of key personnel to conduct

unnecessary testing.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 18, 1995

Description of amendment request: The proposed changes to the

Technical Specifications (TS) would change TS Table 4.3.7.1-1

``Radiation Monitoring Instrumentation Surveillance Requirements.''

This change would increase the channel functional test interval from

monthly to quarterly for each instrument.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

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

consequences of an accident previously evaluated.

The proposed change involves no hardware changes, no changes to

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

existing structures. Increasing the interval between channel

functional tests for the radiation monitoring instrumentation

represent changes that do not affect plant safety and do not alter

existing accident analyses.

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

accident from any previously evaluated.

The proposed change is procedural in nature concerning the

channel functional test frequency for the radiation monitoring

instrumentation not already on a quarterly surveillance. The channel

functional test methodology for these instruments remains unchanged.

The proposed changes, while slightly increasing the possibility of

an undetected instrument error, will not create a new or unevaluated

accident or operating condition.

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

safety.

The proposed change is in accordance with recommendations

provided by the NRC regarding the improvement of Technical

Specifications. These changes will result in perpetuation of current

safety margins while reducing regulatory burden and decreasing

equipment degradation.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: May 4, 1995

Description of amendment request: The proposed change to the

Technical Specifications (TS) would change TS 3/4.6.1.8, ``Drywell and

Suppression Chamber Purge System'', to increase the annual operational

limit for the drywell and suppression chamber purge system from 120 to

500 hours.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

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

consequences of an accident previously evaluated.

The proposed change involves no hardware changes and no changes

to existing structures. Increasing the annual operational limit of

the drywell and suppression chamber purge system will not increase

the probability of a loss-of-coolant accident. While increased usage

of the purge system will result in a slight increase in the

possibility that these valves will be open during a LOCA, it will

not alter or impact previous LOCA analyses.

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

accident from any previously evaluated.

The proposed change will not result in an unanalyzed condition.

While the increase in purge system operation will slightly increase

the possibility of the containment vent and purge valves being open

at the onset of a LOCA event, the valves have been established as

capable of isolating the containment within five seconds. This is

well within the bounds of existing LOCA analyses which assume an

open duration of 175 seconds. Therefore, this change will not

require a new or different accident analysis.

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

safety.

The proposed change will not alter existing systems, equipment,

components, or structures. The method of operating the drywell and

suppression chamber purge system will not be altered by the

increased annual usage. While there is a slight increase in the

possibility of purge operations at the onset of a LOCA, any

resulting release would be insignificant and bounded by existing

LOCA analyses. Operation of the drywell and suppression chamber

purge system based on these proposed changes will remain within the

guidance provided in the NRC's Branch Technical Position CSB 6-4.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Pennsville Public Library,

190 S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: John F. Stolz

Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146,

Saxton Nuclear Experimental Facility (SNEF), Bedford County,

Pennsylvania

Date of amendment request: June 2, 1995, as supplemented on June

23, 1995.

Description of amendment request: The proposed changes to the

technical specifications are administrative in

[[Page 42608]]

nature. The proposed amendment would revise the organization structure

associated with the SNEF to allow General Public Utilities Nuclear

Corporation resources to be applied to SNEC activities within their

normal organizational structure; eliminating the need to identify and

compartmentalize a portion of the organization as specific to SNEC. The

proposed amendment would also revise the description and drawing of the

SNEF site to reflect multiple gates in the SNEF fence.

Basis for proposed no significant hazards

considerationDetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below: The proposed changes

do not involve a significant hazards considerations because the changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The administrative changes will not impact the physical

condition of the containment vessel as it relates to the risk of

fire, flood or radiological hazard.

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

from any previously analyzed.

In its present condition, the only accidents applicable to the

site are those addressed above.

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

The proposed administrative changes would have no effect on any

margins of safety for any evaluated accidents.

The NRC staff has reviewed the analysis of the licensee and, based

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Saxton Community Library, 911

Church Street, Saxton, Pennsylvania 16678Attorney for the Licensee:

Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: Seymour H. Weiss

South Carolina Electric & Gas Company, South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: June 30, 1995

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) for the pressurizer power

operated relief valves (PORVs) to follow the guidance of Generic Letter

(GL) 90-06, Generic Issue 70, and the improved Westinghouse

Standardized Technical Specifications (NUREG-1431, Rev. 1).

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The probability or consequences of an accident previously

evaluated is not significantly increased.

There is no increase in the probability of an accident because

the physical characteristics of the PORVs and their block valves

remain unchanged. No changes to any hardware or software that

affects these components is planned.

The PORVs are pressure relieving devices and only two failure

modes need to be considered. The first is that one or more PORVs or

block valves fail to open when required. This is not

a significant concern and is not a credible cause of any

accident. The second mode is failing to close which includes

depressurization of the RCS [reactor coolant system] and a reactor

trip on low pressurizer pressure or overtemperature [delta]T. The

consequences for the more limiting Pressurizer Safety Valve

Accidental Depressurization event has been analyzed with acceptable

results.

There is no increase in the consequences of an accident as a

result of this change, because only one PORV is required to mitigate

the consequences of a design basis Steam Generator Tube Rupture.

There is sufficient redundancy to ensure one PORV is available to

perform this function even if one PORV is inoperable or incapable of

being manually cycled. The validation of the Emergency Operating

Procedures on the VCSNS [Virgil C. Summer Nuclear Station] simulator

demonstrated that one pressurizer PORV has sufficient capacity to

depressurize the RCS in a time frame which will not cause the

offsite doses presented in the FSAR [Final Safety Analysis Report]

to be exceeded.

The PORVs are utilized to depressurize the RCS and equalize the

pressure between the primary and secondary systems. This stops the

intrusion of RCS water into the secondary which can be released into

the atmosphere. By the time the PORVs are called upon, the affected

steam generator (SG) has been identified and steps have been taken

to isolate the faulted SG. This acts to minimize the radiological

impact on the health and safety of the public. In all cases, the

dose results are within 10 CFR 100 limits.

2. The possibility of an accident or a malfunction of a

different type than any previously evaluated is not created.

The proposed TSCR [TS Change Request] does not involve any

physical changes to the plant or decrease the number of PORVs and

block valves that must be capable of performing their intended

function. These components are used to mitigate the effects of

postulated events and their failure has already been considered. The

worst case failure, either not opening or not closing, has been

evaluated and is bounded by other more limiting accidents.

3. The margin of safety has not been significantly reduced.

The currently approved TS permits all three PORVs and/or their

block valves to be inoperable as long as precautions are taken to

assure that RCS would not leak-by, assuming single failures and

spurious operation. The proposed TSCR would require a minimum of two

PORVs and block valves to be operable, or at least capable of being

manually cycled, in Modes 1, 2, and 3. This is in fact an increase

in margin and provides for greater reliability with the added

benefit that the probability of challenges to the pressurizer code

safety valves will be lessened.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: Frederick J. Hebdon

South Carolina Electric & Gas Company (SCE&G), South Carolina

Public Service Authority, Docket No. 50-395, Virgil C. Summer

Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: July 28, 1995

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to exclude the requirement to

perform the slave relay test of the 36-inch containment purge supply

and exhaust valves on a quarterly basis while the plant is in Modes 1,

2, 3, or 4.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No, the probability or consequences of an accident previously

evaluated would not be increased since no credit is taken for the

valves in FSAR [Final Safety Analysis Report] Chapter 15.

The only credible accident discussed in FSAR Chapter 15 that

applies to these valves is a fuel handling accident inside

[[Page 42609]]

containment (15.4.5.1). The analysis assumes the escaped gases are

released instantaneously to the environment via the Reactor

Building purge system. The analysis does not take credit for

these valves nor for filtration or holdup time during release. The

result of the analysis is acceptable and offsite doses are within

the limits of 10 CFR 100.

TS 3.6.1.7 requires that these valves be sealed shut during

Modes 1, 2, 3, and 4. When sealed shut, these valves will not open

via any signal.

With these valves already in a shut position, neither the

probability nor the consequences of an accident are increased.

2. Does the change create the possibility of a new or different

kind of accident from any previously evaluated?

No, the 36'' [inch] containment purge exhaust and supply valves

will not be placed in a condition different from that evaluated

previously.

The only credible accident discussed in FSAR Chapter 15 that

applies to these valves is a fuel handling accident inside

containment (15.4.5.1). The analysis assumes the escaped gases are

released instantaneously to the environment via the Reactor Building

purge system. The analysis does not take credit for these valves nor

for filtration or holdup time during release. The result of the

analysis is acceptable and offsite doses are within the limits of 10

CFR 100.

Additionally, TS 3.6.1.7. requires that these valves be sealed

shut during Modes 1, 2, 3, and 4. When sealed shut, these valves

will not open via any signal.

3. Does the change involve a significant reduction in the margin

of safety?

TS 4.3.2.1. requires that this slave relay test be performed

quarterly. This surveillance is accomplished for the 36'' [inch]

containment purge exhaust and supply valves by cycling the

respective K615 relay. This will not provide assurance that the

valve will perform its safety function since the valve is sealed

closed. The proposed change will exclude the requirement to perform

the K615 relay test (auto actuation logic and actuation relays -

slave relay test) on a quarterly basis while the plant is in Modes

1, 2, 3,or 4.

TS 3.6.1.7. requires that these valves be sealed shut during

Modes 1, 2, 3, and 4. When sealed shut, these valves will not open

via any signal. Since this relay would not be needed to supply a

signal to place these valves in the closed position, the margin of

safety is not affected.

Based on the preceding analysis, SCE&G has determined that this

change does no involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: June 2, 1995 (TS 353)

Description of amendment request: The proposed amendment supports

replacement of the existing power range neutron monitoring equipment

and implements ARTS/MELLL [average power range monitor and rod block

monitor technical specifications/maximum extended load line limit]

analysis improvements.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Group A Changes: This proposed TS change is associated with the

NUMAC PRNM [nuclear measurement analysis and control power range

neutron monitor] retrofit design. The proposed TS change involves

modification of the LCOs [limiting condition for operations] and SRs

[surveillance requirements] for equipment designed to mitigate

events which result in power increase transients. For the APRM

[average power range monitor] system mitigative action is to block

control rod withdrawal or initiate a reactor scram which terminates

the power increase when setpoints are exceeded. For the RBM [rod-

block monitor] system mitigative action is to block continuous

control rod withdrawal prior to exceeding the MCPR [minimum critical

power ratio] safety limit during a postulated Rod Withdrawal Error

[RWE]. The worst case failure of either the APRM or the RBM systems

is failure to initiate mitigative action (failure to scram or block

rod withdrawal). Failure to initiate mitigative action will not

increase the probability of an accident. Thus, the proposed change

does not increase the probability of an accident previously

evaluated.

For the APRM and the RBM systems, the NUMAC PRNM design,

together with revised operability requirements (LCOs) and revised

testing requirements (SRs), results in equipment which continues to

perform the same mitigation functions under identical conditions

with reliability equal to or greater than the equipment which it

replaces. Because there is no change in mitigation functions and

because reliability of the functions is maintained, the proposed

change does not involve an increase in the consequences of an

accident previously evaluated.

Group B Changes: This proposed change is associated with

implementation of the ARTS/MELLL analysis. The proposed change will

permit expansion of the current allowable power/flow operating

region and will apply a new methodology for assuring that fuel

thermal and mechanical design limits are satisfied. Reference 3

evaluates operation in the MELLL region with assumed implementation

of the ARTS changes. The conclusion of reference 3 is that for all

events and parameters considered there is adequate design margin for

operation in the MELLL region. Because operation in the MELLL region

maintains adequate design margin, the proposed change does not

significantly increase the probability of an accident previously

evaluated.

In support of operation in the MELLL region, the proposed change

modifies flow-biased APRM scram and rod block setpoints and

implements new RBM power-biased setpoints. This potentially changes

the way in which the APRM and RBM systems perform their mitigation

functions. However, no credit for the flow-biased APRM scram or rod

block is taken in mitigation of any design basis event; thus,

changing the APRM setpoints does not impact the consequences of any

accident previously evaluated. The proposed changes to the RBM

system potentially impact mitigation of the RWE. However, per

discussion in reference 3, the proposed RBM changes will assure that

the RWE is not a limiting event; thus, the consequences of the RWE

are not increased. The proposed change does not increase the

consequences of an accident previously evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes (Group A and Group B) involve modification

and replacement of the existing power range neutron monitoring

equipment, modification of the setpoints and operational

requirements for the APRM and RBM systems, implementation of a new

methodology for administering compliance with fuel thermal limits,

and operation in an extended power/flow domain. These proposed

changes do not modify the basic functional requirements of the

affected equipment, create any new system interfaces or

interactions, nor create any new system failure modes or sequence of

events that could lead to an accident. The worst case failure of the

affected equipment is failure to perform a mitigation action, and

failure of this mitigative equipment does not create the possibility

of a new or different kind of accident. The proposed change does not

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

any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

Group A Changes: This proposed TS change is associated with the

NUMAC PRNM retrofit design. The NUMAC PRNM change does not impact

reactor operating parameters nor the functional requirements of the

power

[[Page 42610]]

range neutron monitoring system. The replacement equipment continues to

provide information, enforce control rod blocks and initiate reactor

scrams under appropriate specified conditions. The proposed change

does not revise any safety margin requirements. The replacement

APRM/RBM equipment has improved channel trip accuracy compared to

the current system and meets or exceeds system requirements

previously assumed in setpoint analysis. Thus, the ability of the

new equipment to enforce compliance with margins of safety equals or

exceeds the ability of the equipment which it replaces. The proposed

change does not involve a reduction in a margin of safety.

Group B Changes: This proposed change is associated with

implementation of recommendations presented in the ARTS/MELLL

analysis. Operation in the MELLL region does not affect the ability

of the plant safety-related trips or equipment to perform their

functions, nor does it cause any significant increase in offsite

radiation doses resulting from any analyzed event. Analyses

documented in reference 3 demonstrate that for operation in the

MELLL region adequate margin to design limits is maintained.

Implementation of the ARTS improvements provides flow- and power-

dependent thermal limits which maintain existing margins of safety

in normal operation, anticipated operational occurrences and

accident events. Implementation of power-biased RBM setpoints

improves the margin of safety in a postulated RWE by assuring that

the RWE is not a limiting event. The proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: June 8, 1995 (TS 361)

Description of amendment request: The proposed amendment clarifies

the definition of operability for the RHRSW system standby coolant

supply capability and revises the instrument numbers for several

instruments that have been upgraded.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change to TS 3.5.C.3 clarifies the operability

requirements of the standby coolant supply capability. It does not

change or degrade the nuclear safety characteristics of the RHRSW

and RHR systems and will not affect the intent of the TS. The

operation of the standby coolant supply capability is not a

precursor to any design basis accident or transient analyzed in the

BFN FSAR. The proposed changes to instrument numbers are

administrative changes for the upgraded drywell temperature and

pressure instrumentation. The proposed changes do not affect the

design basis or the safety functions of the Primary Containment

system, since the function and instrumentation range is not changed.

Therefore, the probability of occurrence or the consequences of an

accident or malfunction of equipment important to safety previously

evaluated in the safety analysis report has not been increased.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The possibility for an accident or malfunction of a different

type than any evaluated previously in the safety analysis report is

not created by this change. The change to TS 3.5.C.3 adds the

indication of associated valves of the function involved and a

clarification of operability for the standby coolant supply

connection to be commensurate with the RHR cross-connect capability.

The proposed changes to instrument numbers are administrative

changes effected by the upgrade of instrumentation. There are no

automatic actions affected or compromised by these changes.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change to TS 3.5.C.3 does not affect any acceptable

limit of operation or analysis assumption in the TS or Bases. The

changes affect neither setpoints, calibration intervals, nor

functional test intervals. The change does not affect any acceptable

limit of operation or analysis assumption found in the TS or their

bases. The proposed administrative changes to the instrument numbers

do not affect the setpoint, calibration interval or function of the

instrumentation. These changes do not affect any limiting conditions

of operation or analysis assumption in the TSs or their bases.

Therefore, the change does not reduce the margin of safety as

defined in the basis for any TS.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: June 16, 1995 (TS 360)

Description of amendment request: The proposed change will revise

the BFN Units 1, 2, and 3 Technical Specifications (TS) to permit the

Traversing In-Core Probe (TIP) system to be considered operable with

less than five TIP machines operable. The proposed amendment will allow

the utilization of substitute data in lieu of data from inaccessible

TIP measurement locations. The substitute data will be derived from

either symmetric TIP measurement locations (under certain core

conditions) or from normalized TIP data as calculated by the on-line

core monitoring system.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The TIP system is not used to prevent, or mitigate the

consequences of any previously analyzed accident or transient; nor

are any assumptions made in any accident analysis relative to the

operation of the TIP system. The primary containment isolation

function (TIP withdrawal) is not affected. The

proposed TS change does not alter the fundamental process

involved in calibrating neutron instrumentation (LPRMs) [local power

range monitors], but requires that only the equipment associated

with the TIP channels necessary for recalibrating LPRMs and for core

monitoring functions be operable. Collection and storage of TIP data

without using all TIP channels is acceptable because TIP machine

normalization factors are ultimately derived from the most recent

full core TIP set, which intercalibrates the TIP machines in a

common core location.

Additionally, the use of symmetric detectors and analytical

values as substitute data for inaccessible TIP channels does not

compromise the ability of the process computer to accurately

represent the spatial neutron flux distribution of the reactor core.

[[Page 42611]]

The core monitoring methodology is presently based on symmetry of rod

patterns and fuel loading. This is not changed but extended to use a

higher order of symmetry (octant symmetry) which exists with ``type

A'' sequence rod patterns. Therefore, this change does not increase

the probability or consequences of an accident previously evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed change does not involve the installation of any new

equipment, or the modification of any equipment designed to prevent

or mitigate the consequences of accidents or transients. Therefore,

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

different kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The total core TIP reading uncertainties will remain within the

assumptions of the licensing basis. Therefore, the margin of safety

to the MCPR [minimum critical power ratio] safety limits is not

reduced. The ability of the process computer to accurately represent

the spatial neutron flux distribution for the reactor core is not

compromised. Additionally, the computer's ability to accurately

predict the LHGR [linear heat generation rate], APLHGR [average

planar linear heat generation rate], MCPR and its ability to provide

for LPRM calibration is not compromised. Therefore, the proposed

changes do not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: October 21, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.6.1.2, ``Primary Containment

Leakage.'' The changes would clarify that the main steam line isolation

valves leakage is accounted for separately from the integrated primary

containment leak rate or combined local leak rate results. Also, two

references would be deleted, the test duration for use of Bechtel

Corporation Topical Report BN-TOP-1 would be clarified, and the

requirement to perform the third integrated leak rate in each 10-year

service period in conjunction with the 10-year plant inservice

inspection would be deleted. Exemptions to 10 CFR Part 50 Appendix J,

``Primary Reactor Containment Leakage Testing for Water-Cooled Power

Reactors,'' are also being requested in conjunction with the proposed

TS changes.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration which is presented below:

Part A - Formalize the Approval for Excluding the Main Steam

Line Isolation Valve Leakages from Inclusion in i) the Overall

Integrated Primary Containment Leak Rate and ii) the Combined Local

Leak Rate, and Clarify that the Main Steam Lines are Not Required to

be Vented and Drained for Type A Testing

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

the probability or consequences of an accident previously evaluated.

Since Appendix J was originally envisioned, alternative means of

meeting the intent of these requirements have been developed which

provide an equivalent level of protection of the public health and

safety. However, since some of these alternatives deviate from the

specific wording of Appendix J, exemptions are appropriate for these

alternatives. Implicit in the FSAR treatment of the main steam line

leakage, as well as the TS requirements for main steam line leakage,

are several deviations from the specific requirements of Appendix J.

Although PNPP's methods and practices for Appendix J testing have

been previously described in correspondence to the NRC, a formal

exemption was not recognized to be needed at that time in that the

NRC's approval was perceived to be received by the issuance of the

PNPP TS. Exemption to four separate paragraphs of 10 CFR 50 Appendix

J will document the approvals previously received and incorporated

into the TS for main steam line isolation valve testing during the

initial licensing of the PNPP. This TS change adds references to

footnotes within the TS LCO 3.6.3.1 to clarify which conditions

represent exemptions to Appendix J. These exemptions are described

in the Bases.

PNPP utilized the criteria described in the Standard Review Plan

(SRP), Section 15.6.5, Appendix D, ``Radiological Consequences of a

Design Basis Loss-of-Coolant Accident: Leakage from Main Steam

Isolation Valve Leakage Control System (Rev. 1 - July 1981).'' This

is an alternative, NRC approved method for assessing the MSIV

leakage contribution and determining the radiological consequences.

In accordance with the SRP, the safety analysis for a design

basis LOCA includes the maximum main steam line leak rate separately

from the maximum containment leak rate. Within Appendix J it is

implied that Type A tests are intended to measure the primary

containment overall integrated leak rate, but this vas before the

SRP Section was developed which allows the MSIV contribution to be

accounted for separately in the safety analysis. Therefore, the MSIV

leak rate should not be included in the measurement of the ILRT.

Including the MSIV leakage in the combined local leak rate limit is

also not necessary since a specific Type C MSIV leak rate has been

specified in TS 3.6.1.2.

In summary, there is no change in the probability or

consequences of any accident since the addition of the references

and footnotes to clarify the TS LCO and Actions do not change the

design of the plant, nor the operational characteristics of any

plant system, nor the procedures by which the Operators run the

plant. These changes only cite formal Appendix J exemptions which

are requested to document the approval previously received. A formal

request for exemption to the applicable paragraphs of 10 CFR 50

Appendix J is also being submitted in a separate letter in

conjunction with this proposed TS change.

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

or different kind of accident from any accident previously

evaluated.

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

different kind of accident from any accident previously evaluated.

There are no design changes being made that would create a new type

of accident or malfunction, and the method and manner of plant

operation remains unchanged. The only change being made is an

exemption to 10 CFR 50 Appendix J which will be cited in the TS to

document the implicit and explicit approvals of the PNPP design and

testing methods for main steam line isolation valves. The

requirements and bases for which the formal exemption is sought are

currently presented and implemented in the licensing basis and the

TS for PNPP. The objective of the regulation is being met and will

continue to be met. The exemption to 10 CFR 50 Appendix J is being

submitted in a separate letter in conjunction with this proposed TS

change.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

These changes do not involve a significant reduction in the

margin of safety because they are administrative in nature. The

proposed change will only cite the NRC exemption that grants the

deviation from Appendix J. The proposed changes do not affect any

USAR design bases or accident assumptions. Therefore, the proposed

changes do not reduce the margin of safety as defined in the bases

for any Technical Specification.

Part B - Revise Surveillance Requirement 4.6.1.2 to Eliminate

Unnecessary References and ClarifY the Use of BN-TOP-1

[[Page 42612]]

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

the probability or consequences of an accident previously evaluated.

Surveillance Requirement 4.6.1.2 is proposed to be revised to

eliminate the direct reference to the ANSI Standards N45.4 and N56.8

within the text, because these same Standards are listed within

Appendix J. It is unnecessary to repeat the references to the

Standards within the Technical Specifications because the PNPP is

still required to be in compliance with the regulations. No

additional benefits are gained and licensee flexibility to upgrade

to later versions of the Standards is reduced since a Technical

Specification change is necessary to change the version of the

Standard to which PNPP is committed. This change removes a redundant

requirement to list these Standards in the Technical Specifications.

Therefore, this change cannot involve a significant increase in the

probability or consequences of an accident because the regulation is

still required to be met.

A reference to Topical Report BN-TOP-1 continues to be retained

within Surveillance Requirement 4.6.1.2, and the use of the report

is clarified to be for test durations less than 24 hours. This

reference is retained within the TS since a reference to BN-TOP-1,

though not specifically included within Appendix J, is allowed by

Section 7.6 of ANSI N45.4-1972 and has been approved for PNPP use by

the NRC. The TS Bases are also proposed to be revised to include a

statement that the use of BN-TOP-1 is in accordance with Appendix J.

These changes result in no changes to plant systems and have no

effect on accident conditions or assumptions. These proposed changes

do not affect possible initiating events for accidents previously

evaluated, or any system functional requirements. Hence, these

changes are purely administrative in that they are designed to

eliminate a redundant requirement and clarify the applicability and

acceptability of an alternative leak rate testing provision within

the TS. These changes do not affect plant operation in any way.

Therefore, the proposed changes do not affect the probability or

consequences of any accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

There are no design changes being made that would create a new

type of accident or malfunction, and the method and manner of plant

operation remains unchanged. These changes eliminate a redundant

requirement and clarify the applicability and acceptability of

alternative leak rate testing provisions within the TS. Since the

alternative leak rate testing provisions have been approved by the

NRC, the objective of the regulation continues to be met. Therefore,

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

different kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

These changes do not involve a significant reduction in the

margin of safety because they are administrative in nature and

either eliminate a redundant requirement or clarify the

applicability and acceptability of an alternative, NRC approved,

leak rate testing provision within the TS. The proposed changes do

not affect any USAR design bases or accident assumptions. Therefore,

the proposed changes do not reduce the margin of safety as defined

in the Bases for any Technical Specification.

Part C - Decouple Performance of the Third Type A Test from the

Shutdown for the 10-Year Plant Inservice Inspection

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change revises Surveillance Requirement 4.6.1.2.a

by removing the second sentence requiring that the third test of

each containment Integrated Leak Rate Test (ILRT) set be conducted

during the shutdown for the 10-year plant inservice inspection. A

request for an exemption to 10 CFR 50 Appendix J, Paragraph

III.D.l(a) is also being submitted in conjunction with this proposed

change. Note that this change is also included in the proposed

Appendix J rule changes currently under consideration and has been

approved for several other plants. The deletion of this requirement

from the Technical Specifications does not impact plant safety

because the 10 CFR 50 Appendix J requirement that three Type A

containment ILRT tests to be performed over a 10 year period is not

affected. This change only removes an unnecessary connection between

the two regulations.

The proposed change results in no changes to plant systems. The

proposed change has no effect on accident conditions or assumptions.

The proposed change does not affect possible initiating events for

accidents previously evaluated, or any system functional

requirements. Hence, the proposed change removes an unnecessary tie

between regulations and does not affect plant operation in any way.

In summary, there is no change in the probability or

consequences of any accident since the revision of the existing

Surveillance Requirement to reflect the removal of an unnecessary

tie between regulations does not change the design of the plant, nor

the operational characteristics of any plant system, nor the

procedures by which the Operators run the plant.

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

or different kind of accident from any accident previously

evaluated.

The proposed change removes an unnecessary tie between

regulations. The objective of the regulation continues to be met.

There are no design changes being made that would create a new type

of accident or malfunction, and the method and manner of plant

operation remains unchanged. Therefore, the proposed change does not

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

any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not involve a significant reduction in

the margin of safety because they are administrative in nature and

remove an unnecessary tie between requirements. The proposed change

does not affect any USAR design bases, accident assumptions. or

Technical Specification Bases. Therefore, the proposed change does

not reduce the margin of safety as defined in the bases for any TS.

Based upon the above considerations, it has been concluded that

the proposed changes do not involve significant hazards

considerations.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: June 9 and 30, 1995

Description of amendment request: The licensee has requested a one-

time extension of the performance intervals for certain Technical

Specification Surveillance Requirements (SRs). Affected SRs include

valve testing, and undervoltage instrumentation testing.

Basis for proposed no significant hazards

considerationdetermination: As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed TS change requests one-time only extensions of the

surveillance intervals related to: a) ASME Section XI valve leak

rate, stroke and timing, and position indication testing; b)

Accident Monitoring Instrumentation related to valve position

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

Undervoltage instrumentation LSFT; and, d) leak rate testing for

hydrostatically tested containment isolation valves.

Based on the discussion in the License Amendment Request which

shows:

i) The extension of the interval for ASME Section XI stroke and

timing, leak rate measurement and position indication testing

[[Page 42613]]

requirements are acceptable based on results of past testing which

indicates a margin to TS limits will be maintained;

ii) The extension of the interval for Position Indication

Calibration as specified in Table 4.3.7.5-1, Item 17 is acceptable

based on the testing results from the past two refueling outages

that indicate no failures have occurred:

iii) LSFT interval extension for the Division 1, 2, and 3

Degraded Voltage and Undervoltage instrumentation is acceptable

based on the NRC Safety Evaluation Report (Peach Bottom Atomic Power

Plant, Units 2 and 3, dated August 2, 1993) which supported

extension of the interval for LSFT from 18 to 24 months. This was

based on the small probability of relay or contact failure relative

to mechanical component failure probability and, therefore, the

increase in LSFT interval represented no significant change in the

overall safety system unavailability; and,

iv) The extension of the interval for hydrostatic leak testing

of containment isolation valves is acceptable based on the

consistently low past leak rate data which is a small percentage of

the TS limits.

Therefore, from the above it is shown that the proposed changes

will not significantly increase the probability of an accident

previously evaluated.

The proposed TS change requests one-time only extensions of the

surveillance intervals related to TS SR 4.3.3.1, Table 4.3.3.1-1,

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

Undervoltage instrumentation calibration. [...] extension of the

interval for this instrumentation is acceptable based on the testing

results from the past two refueling outages. No failures have

occurred which would negate the assurance that the instrumentation

would function as required for the requested extended period.

Accordingly, the proposed change will not significantly increase the

probability or consequences of an accident previously evaluated.

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

or

different kind of accident from any accident previously

evaluated.

The proposed TS change requests one-time extensions of the

surveillance intervals for ASME Section XI valve testing,

instrumentation calibration, instrument channel LSFT, containment

isolation valve hydrostatic leak rate testing. The proposed changes

do not necessitate a physical alteration to the plant (no new or

different type of equipment will be installed). In that the

requested extension durations are small as compared to the overall

interval allowed by TS, NRC and industry evaluations support

extension of LSFT, and past testing results provide confidence of no

effect on equipment availability by extending the surveillance

interval, the change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed TS change requests one-time extensions of the

surveillance intervals for the Division 1, 2, and 3 Undervoltage and

Degraded Voltage instrumentation calibration. The proposed changes

do not necessitate a physical alteration to the plant (no new or

different type of equipment will be installed). In that the

requested extension durations are small as compared to the overall

interval allowed by TS and past testing results provide confidence

of no effect on equipment availability by extending the surveillance

interval, the change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change will not involve a significant reduction

in the margin of safety.

The proposed TS change requests a one-time extension of the

surveillance intervals for ASME Section XI valve testing,

instrumentation calibration, instrument channel LSFT, and

containment isolation valve hydrostatic leak rate testing. The

proposed changes do not necessitate a physical alteration to the

plant (no new or different type of equipment will be installed). In

that the requested extension durations are small as compared to the

overall interval allowed by TS, NRC and industry evaluations support

extension of LSFT, and past testing results provide confidence of no

effect on equipment availability by extending the surveillance

interval, the change does not involve a significant reduction in the

margin of safety.

The proposed TS change requests a one-time extension of the

surveillance intervals for the division 1, 2, and 3 Undervoltage and

Degraded Voltage instrumentation calibration. The proposed changes

do not necessitate a physical alteration to the plant (no new or

different type of equipment will be installed). In that the

requested extension durations are small as compared to the overall

interval allowed by TS and past testing results provide confidence

of no effect on equipment availability by extending the surveillance

interval, the change does not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: Gail H. Marcus

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

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

Date of application for amendments: July 19, 1995

Description of amendments request: Amend the Sequoyah Nuclear

Plant, Units 1 and 2 Technical Specification to incorporate new

requirements associated with steam generator tube inspections and

repair.

Date of publication of individual notice in the Federal Register:

August 1, 1995 (60 FR 39198)

Expiration date of individual notice: August 31, 1995

Local Public Document Room Location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these

[[Page 42614]]

amendments. If the Commission has prepared an environmental assessment

under the special circumstances provision in 10 CFR 51.12(b) and has

made a determination based on that assessment, it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

Units 1, 2, and 3, Maricopa County, Arizona

Date of applications for amendments: December 30, 1993 and July 12,

1994. The December 30, 1993, application was supplemented by letters

dated November 30, 1994, May 24, 1995, and June 21, 1995, and the July

12, 1994, application was supplemented by letter dated June 21, 1995.

Brief description of amendments: The amendments (1) revise the

degraded voltage relay trip setpoint and (2) enhance the current

presentation of the information regarding the loss-of-voltage relay

setpoint. A time-voltage curve has been added to the technical

specifications as a more accurate characterization of the inverse-time

relay response.

Date of issuance: July 21, 1995

Effective date: July 21, 1995, to be implemented within 45 days of

issuance.

Amendment Nos.: Unit 1 - Amendment No. 96; Unit 2 - Amendment No.

84; Unit 3 - Amendment No. 67

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 8, 1994 and August

17, 1994 (59 FR 29625 and 59 FR 42334) The November 30, 1994, May 24,

1995, and June 21, 1995, letters provided additional clarifying

information and did not change the initial no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated July 21, 1995.No

significant hazards consideration comments received: No.

Local Public Document Room Location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

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

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

North Carolina

Date of application for amendment: February 6, 1995

Brief description of amendment: The amendment allows the relocation

of cycle-specific core operating limits of Figure 3.1-1, Shutdown

Margin versus Boron Concentration in Technical Specification (TS)

3.1.1.2, Shutdown Margin- Modes 3, 4, and 5, to the plant Core

Operating Limits Report.

Date of issuance: August 1, 1995

Effective date: August 1, 1995

Amendment No. 59

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

Technical Specifications

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

14017) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 1, 1995.No significant

hazards consideration comments received: No

Local Public Document Room Location: Cameron Village Regional

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

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

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

North Carolina

Date of application for amendment: March 30, 1995, as supplemented

July 6, 1995. The July 6, 1995, submittal did not change the initial no

significant hazards consideration determination; it contained

clarifying information only.

Brief description of amendment: The amendment revises the Emergency

Diesel Generator (EDG) surveillance requirements contained in TS 3/

48.1.1.2 to be consistent with NUREG-1431, ``Standard Technical

Specifications, Westinghouse Plants,'' and to eliminate the need for

duplicate EDG testing being performed to satisfy the requirements of

the Station Blackout Rule and the Maintenance Rule.

Date of issuance: August 1, 1995

Effective date: August 1, 1995

Amendment No.: 60

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

Technical Specifications

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20515) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 1, 1995.No significant

hazards consideration comments received: No

Local Public Document Room Location: Cameron Village Regional

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

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

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

Date of application for amendments: June 8, 1995, which superseded

the December 16, 1994, request in its entirety, and additional

correspondence dated November 30, 1994, April 27, May 5, May 11 and

June 23, 1995.

Brief description of amendments: The amendments revised Figure 3.4-

4a ``Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for

the Cold Overpressure Protection (LTOP) System'' in the Braidwood Unit

1's Technical Specifications. The revision extends the applicability of

Figure 3.4-4a from 5.37 effective full power years (EFPY) to 16 EFPY.

In addition, the amendments remove the 638 psig administrative limit

line from the LTOPS curve, because the appropriate instrument

uncertainties and discharge piping pressure limits have been

incorporated in the new curve. Finally, the amendments contains

administrative changes to Figure 3.4-4a and its associated index page.

Date of issuance: July 24, 1995

Effective date: July 24, 1995

Amendment Nos.: 64 and 64

Facility Operating License Nos. NPF-72 and NPF-77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32360). The June 23, 1995, letter, corrected a collating error in the

June 8, 1995, submittal and did not change the initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated July 24, 1995.No significant hazards consideration

comments received: No

Local Public Document Room Location: Wilmington Public Library, 201

S. Kankakee Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of application for amendments: March 23, 1994, as supplemented

on July 26, 1994, and subsequently superseded by a submittal dated

[[Page 42615]]

February 15, 1995. The February 15, 1995, request was supplemented on

February 28, 1995.

Brief description of amendments: The amendments approve a maximum

moderator temperature coefficient (MTC) of +7 pcm/ deg.F and relocate

specification of the cycle specific MTC from the Technical

Specifications to the operating limits report. The staff also approved

the methodology proposed by the licensee for ensuring that the plants

continue to meet the anticipated transient without scram (ATWS) rule

(10 CFR 50.62) during operation with cycle specific MTCs.

Date of issuance: July 27, 1995Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: Byron Units 1 and 2 - 73, 73 and Braidwood Units 1

and 2 - 65, 65

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

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

18623) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 27, 1995.No significant

hazards consideration comments received: No

Local Public Document Room Location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: July 29, 1992, as supplemented

January 14, 1993, February 16, 1993, and May 9, 1995.

Brief description of amendments: The amendments upgrade the current

custom Technical Specifications for Dresden and Quad Cities to the

Standard Technical Specifications contained in NUREG-0123, ``Standard

Technical Specification General Electric Plants BWR/4.'' These

amendments upgrade only Section 3/4.3, ``Reactivity Control.''

Date of issuance: July 27, 1995 Effective date: Immediately, to be

implemented no later than December 31, 1995, for Dresden Station and

June 30, 1996, for Quad Cities Station.

Amendment Nos.: 137, 131, 158, and 154

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 23, 1993 (58 FR

34071) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 27, 1995. No significant

hazards consideration comments received: No

Local Public Document Room Location: for Dresden, Morris Area

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

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

Illinois 61021.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: December 14, 1994

Brief description of amendments: The amendments revise the

surveillance test intervals and allowed outage times for certain

actuation instrumentation in the reactor protection, isolation,

emergency core cooling, control rod withdrawal block, monitoring and

feedwater/main turbine trip systems. The amendments also include

changes to the feedwater/main turbine trip limiting condition for

operation required actions, several mode related changes to the nuclear

instrumentation and rod block specifications, shiftly channel check

requirements for several systems, and several editorial changes to

correct errors and remove outdated footnotes.

Date of issuance: August 2, 1995

Effective date: Immediately, to be implemented within 90 days.

Amendment Nos.: 104 and 90

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: March 1, 1995 (60 FR

11128) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 2, 1995.No significant

hazards consideration comments received: No

Local Public Document Room Location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket No. 50-295, Zion Nuclear Power

Station, Unit 1, Lake County, Illinois

Date of application for amendment: May 17, 1995, as supplemented on

June 2, June 16, and July 12, 1995.

Brief description of amendment: The amendment allows a limited

number of steam generator tubes with roll transition indications to

remain in service until the September 1995 refueling outage.

Date of issuance: July 26, 1995

Effective date: July 26, 1995

Amendment No.: 167

Facility Operating License No. DPR-39: The amendment revises the

Technical Specifications. The June 2, June 16, and July 12, 1995,

submittals provided additional clarifying information that did not

change the initial proposed no significant hazards consideration

determination. The information, however, included changes to details of

the administrative limits mentioned in the initial proposed no

significant hazards consideration determination.Public comments

requested as to proposed no significant hazards consideration

determination: Yes (60 FR 27798). This notice provided an opportunity

to submit comments on the Commission's proposed no significant hazards

consideration determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by June 26, 1995,

but indicated that if the Commission makes a final no significant

hazards consideration determination any such hearing would take place

after issuance of the amendment. The Commission's related evaluation of

the amendments, finding of exigent circumstances and final no

significant hazards consideration determination is contained in a

Safety Evaluation dated July 26, 1995.

Local Public Document Room Location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of application for amendment: December 15, 1994

Brief description of amendment: The amendment revises Technical

Specification 11.3.1.5 ACTION a. to eliminate the need to demonstrate

that the actuation circuitry of the unaffected reactor depressurization

system channels is operable. In addition, the amendment makes an

editorial change to correct a typographical error.

Date of issuance: July 28, 1995

Effective date: July 28, 1995

Amendment No.: 115

Facility Operating License No. DPR-6. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20516) The Commission's related evaluation of the amendment is

contained in a Safety

[[Page 42616]]

Evaluation dated July 28, 1995. No significant hazards consideration

comments received: No.

Local Public Document Room Location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of application for amendment: March 4, 1993, as revised April

14, 1993, as supplemented April 19 and May 31, 1995

Brief description of amendment: The amendment revises the Technical

Specifications (TS) to conform to the wording of the revised 10 CFR

Part 20, ``Standards for Protection Against Radiation,'' and to reflect

a separation of chemistry and radiation protection responsibilities.

Date of issuance: August 2, 1995

Effective date: August 2, 1995

Amendment No.: 16

Facility Operating License No. DPR-6. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 12, 1993 (58 FR

28053), as corrected June 1, 1993 (58 FR 31222). The supplemental

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

Commission's related evaluation of the amendment is contained in a

Safety Evaluation datedNo significant hazards consideration comments

received: No.

Local Public Document Room Location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of application for amendment: April 7, 1994, as

supplementedApril 27, 1995.

Brief description of amendment: This amendment relocates certain

Technical Specifications (TS) that contain fuel cycle-specific

parameter limits that change with core reloads to a Core Operating

Limits Report. TS bases have also been revised to refer to limits

relocated to the COLR. A portion of the amendment request was denied. A

separate Notice of Denial of Amendment has been sent to the Federal

Register for publication.

Date of issuance: July 26, 1995

Effective date: July 26, 1995

Amendment No.: 169

Facility Operating License No. DPR-20. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 25, 1994 (59 FR

27053) The April 27, 1995, submittal provided clarifying information

which was within the scope of the initial application and did not

affect the staff's initial proposed no significant hazards

consideration findings. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated July 26, 1995.No

significant hazards consideration comments received: No.

Local Public Document Room Location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

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

Date of application for amendments: April 12, 1995

Brief description of amendments: The amendments delete Technical

Specification (TS) 3/4.3.4, ``Turbine Overspeed Protection,'' and its

associated Bases. The deletion of TS 3/4.3.4 and its Bases provides

Duke Power Company the flexibility to implement the manufacturer's

recommendations for turbine steam valve surveillance test requirements.

These test requirements will be contained in the Selected Licensee

Commitment Manual.

Date of issuance: July 21, 1995

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

30 days from the date of issuance

Amendment Nos.: 131 and 125

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32361) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 21, 1995. No significant

hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730.

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

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

Date of application for amendments: January 18, 1995.

Brief description of amendments: The amendments relocate the

requirements for the seismic instrumentation, meteorological

instrumentation, and loose-part detection system, and the associated

Bases and surveillance requirements, from the TS to the Selected

Licensee Commitment Manual (Chapter 16 of the FSAR). This will allow

future changes to these controls to be performed under the provisions

of 10 CFR 50.59. No changes are being made to the technical content of

the affected TS pages.

Date of issuance: July 24, 1995

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

30 days

Amendment Nos.: 132 and 126

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 10, 1995 (60 FR

24910) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 24, 1995.No significant

hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: April 12, 1995

Brief description of amendments: The amendments delete Technical

Specification (TS) 3/4.3.4, ``Turbine Overspeed Protection,'' and its

associated Bases. The deletion of TS 3/4.3.4 and its associated Bases

provides Duke Power Company the flexibility to implement the

manufacturer's recommendations for turbine steam valve surveillance

test requirements. These test requirements will be contained in the

Selected Licensee Commitment (SLC) Manual. The SLC Manual is Chapter 16

of the Updated Final Safety Analysis Report.

Date of issuance: August 2, 1995

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

30 days from the date of issuance

Amendment Nos.: 156 and 138

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32362) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 2, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room Location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: September 28, 1994, as

supplemented

[[Page 42617]]

by letters dated May 3 and June 14, 1995.

Brief description of amendments: The amendments revise Technical

Specification Tables 3.3-3, 3.3-4, 3.3-5, and 4.3-2 of the Engineered

Safety Features Actuation System Instrumentation tables to update the

``Loss of Power'' function.

Date of issuance: August 2, 1995

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

60 days, or 60 days after the completion date of the Unit 2

modification, whichever is later.

Amendment Nos.: 157 and 139

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

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

FR 65811) The May 3 and June 14, 1995, letters provided clarifying

information that did not change the scope of the September 28, 1994,

application and the initial proposed no significant hazards

consideration determination. The

This text is long and has been trimmed here. Open the source document for the complete record.