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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 8, 1996, through November 21, 1996.

The last biweekly notice was published on November 19, 1996.

NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENTS TO FACILITY

OPERATING LICENSES, PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION, AND OPPORTUNITY FOR A HEARING

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, 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 January 3, 1997, 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

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effect of any order which may be entered in the proceeding on the

petitioner's interest. The petition should also identify the specific

aspect(s) of the subject matter of the proceeding as to which

petitioner wishes to intervene. Any person who has filed a petition for

leave to intervene or who has been admitted as a party may amend the

petition without requesting leave of the Board up to 15 days prior to

the first prehearing conference scheduled in the proceeding, but such

an amended petition must satisfy the specificity requirements described

above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, 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 20555-0001,

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.

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 amendment request: October 31, 1996

Description of amendment request: The proposed change would revise

the maximum allowable water temperature as measured at the respective

intake structures from 95 deg.F to 94 deg.F and will increase the

minimum main reservoir level from 205.7 feet mean sea level to 215 feet

mean sea level in Technical Specification (TS) 3/4.7.5, Ultimate Heat

Sink.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Since the proposed change does not affect the operation of any

accident initiating systems, the probability of occurrence of an

accident previously evaluated will not increase. Also, none of the

proposed changes will cause plant systems to operate outside their

design limits or create the likelihood of a radioactive release.

Therefore, there would be no increase in 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.

No new component or system level interactions will be created by

the proposed change in ultimate heat sink level and temperature, and

no design limits will be exceeded. This change to [Technical]

Specification 3/4.7.5 is more conservative than the current

Specification limits and will serve only to restrict operation to a

higher reservoir level and lower temperature than was previously

allowed. 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 amendment does not involve a significant

reduction in the margin of safety.

The proposed amendment will establish a more conservative

minimum main reservoir level such that safety-related heat

exchangers served by Emergency Service Water will continue to remove

their design-basis accident heat loads. Establishing a higher

minimum reservoir level, combined with a more conservative reservoir

temperature assumption, will involve an increase in the margin of

safety. Also, the proposed change in maximum reservoir temperature

from 95 deg.F to 94 deg.F will not result in any reduction in the

margin of safety. A maximum pre-accident initial water temperature

of 94 deg.F is necessary to yield a post-accident (30-day)

calculated maximum inlet temperature less than or equal to the

design basis temperature of 95 deg.F. Therefore, the proposed 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.

[[Page 64383]]

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart, Acting

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

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

Date of amendment request: November 4, 1996

Description of amendment request: The proposed amendments would

eliminate from the Technical Specifications, Section 4.7.13.1, the

``during shutdown'' restriction pertaining to the 18-month Standby

Shutdown System (SSS) diesel generator inspection. Unlike Catawba

Nuclear Station, many nuclear plants do not have an SSS facility and

associated diesel generator. The requirements in the Technical

Specifications for the SSS diesel generator (shared between both units)

were patterned after similar requirements for the emergency diesel

generators. The current wording requires that both units be shut down

to perform the subject inspection.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

... The standard for determining that a Technical Specification

amendment request involves no significant hazards considerations

requires that operation of the facility in accordance with the

requested amendment will 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 the margin of safety.

Criterion 1

The proposed amendment seeks to change the surveillance

requirements to allow the SSS DG [diesel generator] periodic

inspection with one or both units on line. The surveillance can be

safely completed as proposed without affecting unit operation. The

equipment would not be removed from service for a time that would

exceed the current Limiting Condition For Operation or the

appropriate action statement would be entered. The probability or

consequences of any accident previously evaluated will not be

significantly increased because the removal of the SSS DG from

service can be safely performed while one or both units are

operating.

Criterion 2

The proposed amendment change does not change any actual

surveillance requirements. The change simply allows the 18 month SSS

DG inspection to be performed at different unit conditions. The

performance of the surveillance with the units operating do not

require any new component configurations that would reduce the

ability of any equipment to mitigate an accident. The station is not

degraded beyond that which has been previously evaluated. Therefore

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

different kind of accident.

Criterion 3

The allowed outage time for the SSS DG, as specified by the

Limiting Condition For Operation, defines the required margin of

safety for equipment operability. Removing the SSS DG from service

for periodic inspection and returning it to service within the

allowed outage time 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

proposed amendments involve 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: October 30, 1996

Description of amendment request: The proposed changes would (1)

completely rewrite Technical Specification (TS) 4.4.2 to incorporate a

prestressed concrete containment surveillance program that is

consistent with Regulatory Guide 1.35, (2) modify TS 3.6.7 by

establishing new Limiting Conditions for Operation and required actions

related to the structural integrity of the reactor buildings, (3)

incorporate an editorial change to TS 6.6.3 to reference the relocated

tendon surveillance reporting requirements, and (4) modify TS 3.6.7

Bases to describe the Reactor Building post-tensioning TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed amendment to Oconee Technical Specifications

involves the implementation of an enhanced surveillance program for

the reactor building prestressed concrete containment and the

assurance of appropriate station response to abnormal degradation of

the containment structure. The proposed change will move Oconee into

a surveillance program which is consistent with accepted industry

practice and a published NRC regulatory position. The adoption of

Regulatory Guide 1.35 as a basis for the periodic inspection of the

reactor building prestressed concrete containment and clearly

defined station response to any indication of structural

deterioration will assure acquisition of sufficient data to

demonstrate that structural integrity is maintained and, if

necessary, appropriate compensatory action(s) are taken. By assuring

that any adverse trends in the behavior of the prestressed concrete

containment are identified and acted upon in a timely manner, this

change does not increase the probability or consequences of an

accident previously evaluated.

2) Will the change create the possibility of a new or different

kind of accident from any previously evaluated?

No. The proposed amendment to Oconee Technical Specifications

involves the implementation of an enhanced surveillance program for

the reactor building prestressed concrete containment and the

assurance of appropriate station response to abnormal degradation of

the containment structure. By adopting Regulatory Guide 1.35 as a

basis for the surveillance inspection program for the reactor

building prestressed concrete containment and clearly defining

required station response to any indication of structural

deterioration, sufficient data will be obtained to demonstrate that

structural integrity is being maintained and that any adverse

behavioral trends are identified and acted upon in a timely manner.

Therefore, the proposed amendment does not create the possibility of

any type of accident: new, different or previously evaluated.

3) Will the change involve a significant reduction in a margin

of safety?

No. Margin of safety is associated with confidence in the

ability of the fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure boundary, and containment

structure) to limit the level of radiation dose to the public. The

proposed Technical Specifications amendment will move Oconee into a

surveillance program which is consistent with accepted industry

practice and a published regulatory position. By ensuring more

timely identification of, and response to, any adverse trend in the

behavior of the reactor building prestressed concrete containment,

continued maintenance of the structural integrity is enhanced.

Therefore, the ability of the containment structure to perform the

intended function of protecting the public

[[Page 64384]]

from radiation dose is further assured, and no reduction in any

existing margin of safety will occur.

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: September 9, 1996

Description of amendment request: The proposed amendment would

modify the design features section (Section 5.0) of the Technical

Specifications (TSs) to make the design features section consistent

with the four criteria specified in the Commission's Policy Statement

on TSs (58 FR 39132) and with the guidance provided in the NRC's

Standard Technical Specifications, Westinghouse Plants (NUREG-1431,

Revision 1).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change reduces the content of the technical

specification (TS) design feature section consistent with the

Improved Standard Technical Specifications (ISTS) of NUREG-1431. The

information that has been removed is also contained in the UFSAR

[Updated Final Safety Analysis Report] or offsite dose calculation

manual (ODCM); therefore, duplication of the information is

eliminated to improve the use of the TS. Because the information

removed from the TS is maintained in the UFSAR or ODCM where changes

are controlled in accordance with regulatory requirements, there is

no reduction in commitment and adequate control is provided.

Elimination of information from the design feature section of the TS

which duplicates information in the UFSAR enhances the usability of

the TS without reducing commitments. These changes clarify and

improve the understanding and readability of the TS. 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.

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

kind of accident from any accident previously evaluated?

The relocation of existing requirements, the elimination of

requirements which duplicate existing information, and making

administrative improvements are all changes that are administrative

in nature. The proposed changes will not affect any plant system or

structure, not [nor] will they affect any system functional or

operability requirements. Consequently, no new failure modes are

introduced as a result of the proposed changes. The proposed changes

are consistent with the ISTS, for the most part, as plant-specific

information is included in this section. Therefore, the proposed

change will not create the possibility of a new or different type of

accident from any accident previously evaluated.

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

of safety?

The proposed changes are administrative in nature in that no

change to the design features of the facility are being made. The

design features section is being reformatted to be consistent, for

the most part, with the ISTS. The proposed changes do not affect the

UFSAR design bases, accident assumptions, or technical specification

bases. In addition, the proposed changes do not affect release

limits, monitoring equipment or practices. Therefore, the proposed

change will not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 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

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station,

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 24, 1996

Description of amendment request: The proposed amendment would

revise the technical specifications to remove accelerated testing

requirements for the standby diesel generators. The changes implement

the provisions of Generic Letter (GL) 94-01, ``Removal of Accelerated

Testing and Special Reporting Requirements For Diesel Generators'',

dated May 31, 1994.

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. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

This change will provide flexibility to structure the standby

diesel generator maintenance program based on the risk significance

of the structures, systems, and components that are within the scope

of the Maintenance Rule. The removal of the diesel generator

accelerated testing is acceptable as the maintenance rule applies

site and system specific performance criteria to monitor diesel

generator performance. This criteria includes a running availability

and reliability goal as well as specific goals to monitor

maintenance preventable functional failures. The performance

criteria for the diesel generator reliability and unavailability

established by the maintenance rule and the causal determinations

and corrective actions required for maintenance preventable

functional failures are considered to be an acceptable method for

monitoring diesel generator performance.

The proposed change has no effect on the probability of the

initiation of an accident, because the emergency diesel generators

do not serve as the initiator of any event. Additionally, as diesel

generator performance will continue to be assured by the maintenance

rule, the proposed changes do not affect the ability to mitigate the

consequences of an accident previously evaluated. The changes do not

impact the diesel's design sources, operating characteristics,

system functions, or system interrelationships. The failure

mechanisms for the accidents previously analyzed are not affected

and no additional failure modes are created that could cause an

accident that has been previously evaluated. Since the diesel

generator performance and reliability will continue to be assured by

the maintenance rule, the proposed changes cannot involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. This request does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This proposed change does not involve a change to the plant

design or operation. As a result, the proposed changes does not

affect any of the parameters or conditions that could contribute to

the initiation of any accidents. The proposed changes only affect

the methods used to monitor and assure diesel generator performance.

The performance criteria for both the diesel generator reliability

and unavailability established by the maintenance rule, and the

[[Page 64385]]

casual determinations and corrective actions required for

maintenance preventable functional failures, is considered by GL 94-

01 to be an acceptable method for monitoring diesel generator

performance.

No [system, structure, or component] SSC, method of operating,

or system interface is altered by this change. The changes do not

impact the diesel's design sources, operating characteristics,

system functions, or system interrelationships. The failure

mechanisms for the accidents are not affected, and no additional

failure modes are created. Because the diesel generator performance

and reliability will continue to be assured by the maintenance rule,

the proposed changes cannot create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The request does not involve a significant reduction in a

margin to safety.

The proposed changes only affect the methods used to monitor and

assure diesel generator performance and reliability. The performance

criteria for both the diesel generator reliability and

unavailability established by the maintenance rule, and the casual

determinations and corrective actions required for maintenance

preventable functional failures, is considered by GL 94-01 to be an

acceptable method for monitoring diesel generator performance. No

margin to safety as defined in the basis for any technical

specification is impacted by these changes. This change does not

impact any uncertainty in the design, construction, or operation of

any SSC. Diesel generator response to accident initiators is

unchanged. No SSC, method of operating, or system interface is

altered by this change. The changes do not impact the diesel's

design sources, operating characteristics, system functions, or

system interrelationships. Because the diesel generator performance

and reliability will continue to be assured by the maintenance rule,

the proposed changes cannot involve a significant reduction in the

margin to safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station,

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: November 6, 1996

Description of amendment request: The proposed amendment would

revise the River Bend Station (RBS) Fire Hazards Analysis Report and

Safety Analysis Report to allow a deviation from 10 CFR Part 50,

Appendix R, Section III.G.2.c with respect to the requirement for an

area wide automatic fire suppression system in Fire Area C-16. The

deviation would allow a 1-hour barrier to separate redundant trains of

post fire safe shutdown equipment within the fire area and partial

sprays on the protected train.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The request does not involve an increase in the probability

or consequences of an accident previously evaluated.

The event of concern is a fire in Fire Area C-16. The low fire

loading and minimal concentration of exposed combustible material in

Fire Area C-16 would limit fire spread. However, for this scenario,

all unprotected equipment in Fire Area C-16 will be assumed lost.

Fire Area C-16 contains cables for both Division I and Division II

components required for post fire safe shutdown. The loss of both

divisions of cables could preclude the ability of the plant to

achieve post fire safe shutdown. Protection of the required Division

II cables in a 1-hour fire barrier in conjunction with a partial

area, automatic suppression system installed above and below the

protected trays will ensure that post fire safe shutdown can be

achieved.

In summary, the probability of a fire occurring in Fire Area C-

16 is not affected. However, if a fire were to occur in Fire Area C-

16 which caused the loss of Division I powered components, Division

II powered components, by virtue of the 1-hour fire barrier and

partial area, automatic suppression system, would remain available.

The low fire loading and minimal concentration of exposed

combustible material in Fire Area C-16 would limit fire spread. The

proposed fire protection scheme provides a level of protection

commensurate with the original design. Therefore, this request does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

Fire Area C-16 will be protected by a partial area, automatic

suppression system installed above and below the protected cable

trays. Fire suppression systems are generally used to limit fire

spread, once the heat of the fire opens thermally sensitive

sprinklers. The low fire loading and minimal concentration of

exposed combustible material in Fire Area C-16 would aid in limiting

fire spread, and would also limit the severity of any plausible

fire. The previous analysis assumed all Division I components and

cables in the area would be lost, and that the installed fire

barrier would adequately protect the Division II cables routed

through C-16. The required Division II cables will be enclosed in a

1-hour fire barrier with a partial area, automatic suppression

system. These features provide a level of protection commensurate

with that of the previous design. In addition, the total combustible

loading in the area results in a maximum theoretical worst case fire

duration of less than 1-hour.

In summary, if a fire were to occur in Fire Area C-16 which

caused the loss of Division I powered components, post fire safe

shutdown could still be achieved using Division II. Therefore, this

request does not create the possibility of occurrence of a new or

different kind of accident from any accident previously evaluated.

3. The request does not involve a significant reduction in a

margin of safety.

In this case, the margin of safety is implicit rather than being

explicitly expressed as a numerical value. An implicit margin of

safety involves conditions for NRC acceptance. Since the RBS

Technical Specification Bases do not specifically address a margin

of safety for fire protection, the SAR [Safety Analysis Report], the

NRC's Safety Evaluation Report (SER), and appropriate other

licensing basis documents were reviewed to determine if the proposed

change would result in a reduction in a margin of safety. As stated,

in part, in Attachment 4 to NPF-47 [Facility Operating License; NPF-

47]:

EOI [Entergy Operations, Inc.] shall implement and maintain in

effect all provisions of the approved fire protection program as

described in the Final Safety Analysis Report for the facility

through Amendment 22 and as approved in the SER dated May 1984 and

Supplement 3 dated August 1985 subject to provisions 2 and 3 ....

As discussed in the Reason for Request, SSER [Supplemental

Safety Evaluation Report] 3 dated August 1985 states, in part:

On the basis of its evaluation the staff finds that the

applicant's fire protection program with approved deviations is in

conformance with the guidelines of BTP CMEB [branch technical

position, Chemical Materials and Engineering Branch] 9.5-1,

[S]sections III.G, III.J, and III.O of Appendix R to 10 CFR [Part]

50, and GDC [General Design Criteria] 3, and is, therefore,

acceptable.

Thus, the margin of safety in this case can be defined as

conformance with the specified fire protection guidelines.

10 CFR [Part] 50, Appendix R, Section III.G.2, requires, in

part, that redundant trains of post fire safe shutdown equipment

located in the same fire area be separated by a 1-hour fire barrier

and, in addition, that fire detection and an automatic fire

suppression system be installed in the are under consideration.

Since Fire Area C-16 will have a partial area, automatic suppression

system, this fire area would deviate from the requirements of 10 CFR

[Part] 50, Appendix R, Section III.G.2.c. However, as discussed

previously, the installed partial area, automatic suppression

system, the low fire loading and minimal amount of exposed

combustibles compensate for the lack of a total, area wide,

automatic fire suppression

[[Page 64386]]

system. There is no adverse impact on the ability to achieve and

maintain post fire safe shutdown. Therefore, this request does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant Units 1 and 2, St. Lucie County, Florida

Dates of amendment requests: October 28, 1996 (Two letters)

Description of amendment request: The licensee proposed to change

the St. Lucie Units 1 and 2 Technical Specifications (TS) to implement

10 CFR 50, Appendix J, Option B, for containment leakage testing by

referring to Regulatory Guide 1.163, ``Performance-Based Containment

Leakage-Test Program.'' Changes include relocating the details for

containment testing to the ``containment leakage rate testing program''

and adding the requirements of the containment leakage rate testing

program to TS 6.8.4, which describes facility programs. Changes are

also proposed to remove Tables 3.6-1, ``Containment Leakage Paths,''

and 3.6-2, ``Containment Isolation Valves'' from TS and relocate the

information to plant procedures.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

due to the following reasons:

a)These proposed changes are all consistent with NRC

requirements and guidance for implementation of 10 CFR 50, Appendix

J, Option B, except for the use of Bechtel Topical Report BN-TOP-1

for type A testing. BN-TOP-1 has been previously approved for use in

accordance with 10 CFR 50 appendix J.

b) Based on industry and NRC evaluations performed in support of

developing Option B, these changes potentially result in a minor

increase in the consequences of an accident previously evaluated due

to the increased testing intervals. However, the proposed changes do

not result in an increase in the core damage frequency since the

containment system is used for mitigation purposes only.

c) These changes are expected to result in increased attention

to components with poor leakage test history as part of the

performance-based nature of Option B, such that the marginally

increased consequences from the expanded testing intervals may be

further reduced or negated.

Therefore, these changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.(2) Operation of the facility in accordance with the

proposed amendments would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the implementation of a performance-based program for containment

leakage rate testing, since the proposed changes do not involve the

addition or modification of equipment, nor do they alter the design

or operation of affected plant systems, structures, or

components.(3) Operation of the facility in accordance with the

proposed amendments would not involve a significant reduction in a

margin of safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are basically unchanged by the

proposed amendments. The increase in intervals between leak-test

surveillances will not significantly reduce the margin of safety as

shown by findings in NUREG 1493, ``Performance-Based Containment

Leak-Test Program'', which was based on implementation of the

performance-based testing of Option B.

Therefore these changes do not involve a significant reduction

in the margin of safety.The NRC staff has reviewed the licensee's

analysis and, based on thisreview, 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light,

11770 US Highway 1, North Palm Beach, Fl 33408

NRC Project Director: Frederick J. Hebdon

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389,St. Lucie Plant Units 1 and 2, St. Lucie County, Florida

Date of amendment request: October 30, 1996

Description of amendment request: The proposed amendments will

revise Technical Specification (TS) 3/4.9.10, ``Refueling Operations,

Water Level-Reactor Vessel.'' The Limiting Condition for Operation

(LCO) specified for the minimum allowed refueling water level is not

altered, but the Applicability, Action, and Surveillance Requirements

are changed to remove inconsistencies with the definition of Core

Alterations, and to achieve consistency with the generic Standard

Technical Specifications for Combustion Engineering Plants (NUREG-

1432). An editorial change is proposed for TS 3/4.9.9, ``Refueling

Operations, Containment Isolation System,'' and, for St. Lucie Unit 1,

the LCO is modified to conform with other related refueling

specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Certain evolutions performed with the UGS [upper guide

structure] in place are not Core Alterations, and the revised LCO 3/

4.9.10 will allow these activities to be performed at water levels

other than prescribed by the existing LCO. Since these activities

are performed with the UGS in place, the probability that a fuel

handling accident would occur is not impacted by the proposed

changes. The minimum water level required for Core Alterations and

movement of irradiated fuel in containment is not altered by the

proposed changes, nor are any assumptions or conditions changed that

were used as inputs to the evaluation of fuel handling accident

consequences. The changes to Specification 3/4.9.9 are

administrative in nature and resolve an inconsistency between the

operability requirements for the containment isolation system and

the containment penetrations that the system would isolate at PSL1

[Plant St. Lucie 1]. Therefore, operation of either facility in

accordance with its proposed amendment would not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

amendment would not create the possibility of a new or different

[[Page 64387]]

kind of accident from any accident previously evaluated.

The proposed changes are administrative in nature, in that the

changes do not involve the addition or modification of equipment nor

do they alter the design of plant systems. New failure modes are not

introduced, and the physical plant or the modes of plant operation

defined in the Facility License are not altered. Therefore,

operation of either facility in accordance with its proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The safety margin associated with a fuel handling accident is

determined, in part, by the minimum refueling water level allowed

for conducting Core Alterations and movement of irradiated fuel in

containment. The minimum water level required by LCO 3/4.9.10, or

other factors considered as inputs to the safety analysis, is not

changed by the proposed amendments. The revised applicability

requirements for LCO 3/4.9.9 at PSL1 will allow the containment

isolation system to be inoperable only during those Mode 6

conditions where Core Alterations or irradiated fuel movements

within containment are not in progress, or each required containment

penetration is otherwise closed. Therefore, operation of either

facility in accordance with its 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

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

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

amendmentrequest involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light,

11770 US Highway 1, North Palm Beach, Fl 33408

NRC Project Director: Frederick J. Hebdon

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

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

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: October 23, 1996, as supplemented by

letter dated November 6, 1996.

Description of amendment request: The proposed amendment would

revise Technical Specification 3.4.6.1, regarding reactor coolant

system leakage detection instrumentation, to adopt the requirements

found in NUREG-1431, ``Standard Technical Specifications Westinghouse

Plants,'' for the reactor coolant system leakage detection

instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involved a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change reduces the number of containment

atmospheric radioactivity channels which must be OPERABLE when

operating in MODES 1, 2, 3, and 4 from two to one. This change does

not significantly increase the probability or consequences of a

previously evaluated accident since the plant will continue to have

diverse and independent means of detecting significant changes in

the amount of leakage from the RCS [reactor coolant system]; the

normal sump level and flow monitoring system, at least one of the

two containment atmospheric radiation monitors, and the periodic

precision RCS water inventory balance required by Technical

Specification surveillance requirement 4.4.6.2.1.c. In addition, STP

[South Texas Project] design includes advanced trending displays

which can assist in detecting leakage based on changes in the volume

control tank or pressurizer level. Other instruments, which are not

listed in the Technical Specification related to leakage, but which

can provide indication of leakage, are the containment pressure,

temperature and humidity indicators. Good operating practice and

commercial risk associated with long term inoperability of both

monitors assures that an inoperable containment atmospheric

radiation monitor will be promptly returned to service.

The proposed change also revises the limitation on continued

operation with both containment atmospheric radiation monitors

inoperable from 72 hours to 30 days. This change is based on the

continued availability of diverse and redundant instrumentation

discussed above to detect and indicate RCS leakage.

The Actions required as a result of this change include analysis

of a containment atmospheric grab sample or performance of a

precision RCS water inventory balance in accordance with

surveillance requirement 4.4.6.2.1.c. The containment normal sump

level flow monitoring system will also promptly identify changes in

RCS leakage. Other installed instrumentation, such as containment

pressure, temperature, and humidity, will provide indications of

significant increases in leakage. Slower increases will be detected

by the daily inventory balance or the daily grab samples analysis,

and the three day inventory balance.

Inoperability of the on-line automatic containment normal sump

level and flow monitoring system can be compensated for by the

performance of a daily manual calculation, a precision RCS inventory

balance as described in surveillance requirement 4.4.6.2.1.c, or the

other available indications of increases in leakage such as the

containment atmospheric radiation monitoring instruments and

installed containment temperature, pressure and humidity

instrumentation. The STP control room design also incorporates

features which allow rapid detection of unexpected changes in the

volume control tank and pressurizer level through available

instrument trend displays. The combination of the compensatory

measures, diverse and separate channels, and non-TS [non-technical

specification] required instrumentation provides a sufficient level

of detection to assure prompt identification and quantification of

leakage with an inoperable containment normal sump level and flow

monitoring system. The allowable outage time of 30 days provides

assurance the normal containment sump level and flow monitoring

system will be returned to service in a reasonable amount of time.

Based on the continued availability of adequate and redundant

instrumentation to detect changes in RCS leakage rate, this change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not require the installation of any new

or different kind of equipment. Nor does the change involve any

significant new or different MODE of operation of the plant. The

proposed change reduces the number of required containment

atmospheric radiation monitors, and provides a 30 day allowed outage

time for either the containment atmosphere radioactivity monitor or

the containment normal sump level and flow monitoring system.

Therefore, this 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.

In addition, as described above, the proposed change does not

significantly reduce a margin of safety. Small changes in RCS leak

rates are typically detected over a relatively long period of time.

Diverse instrumentation continues to be available to plant operators

which will assist in early detection of any change. The STP design

provides additional non-Technical Specification human factors which

assist in assuring any changes in leakage will be quickly detected.

The proposed change extends the amount of time that the

containment atmospheric radiation monitors may be inoperable. The

extension is based on the continued availability of equipment which

provides a level of detection capability adequate to detect

increases in RCS leakage and which continues to be diverse and

independent. This protection is afforded by the continued

OPERABILITY of the containment normal sump level and flow monitoring

system, the daily performance of a precision RCS

[[Page 64388]]

inventory balance as described by surveillance requirement

4.4.6.2.1.c or the daily analysis of containment atmospheric grab

samples, and other instrumentation such as pressure, temperature and

humidity indicators.

The combination of the compensatory measures, diverse and

separate channels, and non-TS required instrumentation provides a

sufficient level of detection to assure prompt identification and

quantification of leakage with an inoperable containment normal sump

level and flow monitoring system. Additionally, the compensatory

measure of performing either a daily manual calculation or precision

RCS inventory balance, provides assurance that the level of safety

is maintained when the containment normal sump level and flow

monitoring system is inoperable. The allowable outage time of 30

days provides assurance the normal containment sump level and flow

monitoring system will be returned to service in a reasonable amount

of time.

Based on the compensatory actions and available installed

equipment, the proposed changes do not represent 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869

NRC Project Director: William D. Beckner

Northern States Power Company, Docket Nos. 50-282 and 50-306,

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota

Date of amendment requests: August 15, 1996

Description of amendment requests: The proposed amendments would

revise the Containment Cooling Systems Limiting Conditions for

Operation Technical Specifications to bring them into conformance with

recently completed system analyses by no longer permitting both

containment spray pumps to be inoperable at the same time.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Operation of the Prairie Island plant in accordance with the

proposed changes does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

None of the proposed changes involve a physical modification to the

plant.

These changes will require operability of at least one

containment spray pump at all times and reduces the spray additive

tank allowable outage time from 72 hours to 24 hours. Both of these

changes are more conservative and safer than currently required in

the Prairie Island Technical Specifications. These proposed changes

do allow one containment fan cooler train out of service for 7 days

instead of 72 hours as allowed by current Technical Specifications.

Recent plant analyses confirm that one containment fan cooler train

with one containment spray train is sufficient to meet the system

design bases. Since the probability of an accident occurring is low

while one containment fan cooler train is out of service, the

probability and consequences of an accident are not significantly

increased.

In total these changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

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

different kind of accident from any accident previously evaluated

because the proposed changes, in themselves, do not introduce a new

mode of plant operation, surveillance requirement or involve a

physical modification to the plant.

The proposed changes do require more restrictive, safer

containment spray train operability. The proposed changes also allow

one containment fan cooler train to be out of service for 7 days

instead of 72 hours as allowed by the current Technical

Specifications. However, this change does not create the possibility

of a new kind of accident.

The proposed changes do no alter the design, function, or

operation of any plant components and therefore, no new accident

scenarios are created.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be created

by these amendments.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety. This License Amendment Request

require[s] one containment spray train to be operable at all times

which is more restrictive than current Technical Specifications and

thus the margin of safety is not reduced.

This License Amendment Request will also allow one containment

fan cooler train to be out of service for 7 days instead of 72 hours

as allowed by the current Technical Specifications. Since the

remaining containment cooling components can mitigate an accident

and the probability of a design basis accident are low during this

time, this change does not significantly reduce the plant margin of

safety.

Therefore, a significant reduction in the margin of safety would

not be involved with these amendments.

Based on the evaluation described above, and pursuant to 10 CFR

Part 50, Section 50.91, Northern States Power Company has determined

that operation [of] the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by Nuclear

Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.

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: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

Northern States Power Company, Docket Nos. 50-282 and 50-306,

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota

Date of amendment requests: September 24, 1996, as supplemented

October 17, 1996.

Description of amendment requests: The proposed amendments would

revise the Technical Specifications (TS) for the Prairie Island Nuclear

Generating Plant to allow use of an alternate steam generator tube

repair criteria (elevated F-star or EF*) in the tubesheet region when

used with the repair method of additional roll expansion. The

amendments incorporate revised acceptance criteria for tubes with

degradation in the tubesheet region and enable the licensee to avoid

unnecessary plugging and sleeving of steam generator tubes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The supporting technical and safety evaluations of the subject

criterion

[[Page 64389]]

demonstrate that the presence of the tubesheet will enhance the tube

integrity in the region of the hardroll by precluding tube

deformation beyond its initial expanded outside diameter. The

resistance to both tube rupture and tube collapse is strengthened by

the presence of the tubesheet in that region. The results of

hardrolling of the tube into the tubesheet is an interference fit

between the tube and the tubesheet. Tube rupture cannot occur

because the contact between the tube and tubesheet does not permit

sufficient movement of tube material. The radial preload developed

by the rolling process will secure a postulated separated tube end

within the tubesheet during all plant conditions. In a similar

manner, the tubesheet does not permit sufficient movement of tube

material to permit buckling collapse of the tube during postulated

LOCA [loss-of-coolant accident] loadings.

The EF* length of roll expansion is sufficient to preclude tube

pullout from tube degradation located below the EF* distance,

regardless of the extent of the tube degradation. The existing

Technical Specification leakage rate requirements and accident

analysis assumptions remain unchanged in the unlikely event that

significant leakage from this region does occur. As noted above,

tube rupture and pullout is not expected for tubes using the EF*

criterion. Any leakage out of the tube from within the tubesheet at

any elevation in the tubesheet is fully bounded by the existing

steam generator tube rupture analysis included in the Prairie Island

Plant USAR [updated safety analysis report]. For plants with partial

depth roll expansion like Prairie Island, a postulated tube

separation within the tube near the top of the roll expansion (with

subsequent limited tube axial displacement) would not be expected to

result in coolant release rates equal to those assumed in the USAR

for a steam generator tube rupture event due to the limited gap

between the tube and tubesheet. The proposed plugging criterion does

not adversely impact any other previously evaluated design basis

accident.

Leakage testing of roll expanded tubes indicates that for roll

lengths approximately equal to the EF* distance, any postulated

faulted condition primary to secondary leakage from EF* tubes would

be insignificant.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

Implementation of the proposed EF* criterion does not introduce

any significant changes to the plant design basis. Use of the

criterion does not provide a mechanism to initiate an accident

outside of the region of the expanded portion of the tube. Any

hypothetical accident as a result of any tube degradation in the

expanded portion of the tube would be bounded by the existing tube

rupture accident analysis. Tube bundle structural integrity will be

maintained. Tube bundle leaktightness will be maintained such that

any postulated accident leakage from EF* tubes will be negligible

with regard to offsite doses.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The use of the EF* criterion has been demonstrated to maintain

the integrity of the tube bundle commensurate with the requirements

of Reg Guide 1.121 [Bases for Plugging Degraded PWR Steam

Generator Tubes] (intended for indications in the free

span of tubes) and the primary to secondary pressure boundary under

normal and postulated accident conditions. Acceptable tube

degradation for the EF* criterion is any degradation indication in

the tubesheet region, more than the EF* distance below the bottom of

the transition between the roll expansion and the unexpanded tube.

The safety factors used in the verification of the strength of the

degraded tube are consistent with the safety factors in the ASME

[American Society of Mechanical Engineers] Boiler and Pressure

Vessel Code used in steam generator design. The EF* distance has

been verified by testing to be greater than the length of roll

expansion required to preclude both tube pullout and significant

leakage during normal and postulated accident conditions. Resistance

to tube pullout is based upon the primary to secondary pressure

differential as it acts on the surface area of the tube, which

includes the tube wall cross-section, in addition to the inner

diameter based area of the tube. The leak testing acceptance

criteria are based on the primary to secondary leakage limit in the

Technical Specifications and the leakage assumptions used in the

USAR accident analyses.

Implementation of the tubesheet plugging criterion will decrease

the number of tubes which must be taken out of service with tube

plugs or repaired with sleeves. Both plugs and sleeves reduce the

RCS (reactor coolant system) flow margin; thus, implementation of

the EF* criterion will maintain the margin of flow that would

otherwise be reduced in the event of increased plugging or sleeving.

Based on the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the USAR or the Technical Specification

Bases.

Based on the evaluation described above, and pursuant to 10 CFR

Part 50, Section 50.91, Northern States Power Company has determined

that operation of the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by NRC

regulations in 10 CFR Part 50, Section 50.92.

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: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: June 10, 1996, as supplemented July 25,

1996

Description of amendment request: The proposed amendment would

change the differential temperature Technical Specification Allowable

Values and Trip Setpoints for the Reactor Water Cleanup penetration

room steam leak detection function.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability [of occurrence] [sic] or consequences of an

accident evaluated.

FSAR section 5.2.5.1.3 addresses the ambient and differential

room ventilation temperature leakage detection. This section states:

``...switch setpoints are based on the temperature rise

resulting from a leak at system conditions corresponding to full

reactor power.''

NRC Safety Evaluation on the RWCU system steam leak detection

system (related to Amendment Number 123 to License NPF-14 and

Amendment Number 90 to License NPF-22) reviewed and found acceptable

the PP&L criteria for calculating the leak detection setpoints for

the RWCU system, which include:

1. Setpoints are selected to detect and isolate a leak that is

normally less than 25 gpm and below the flow rate corresponding for

the critical crack size for the system piping.

2. Setpoints are set high enough to avoid inadvertent isolation

caused by normal temperature transients or abnormal transients

caused by non-leak conditions (such as loss of ventilation).

This NRC SER also stated that a leak rate of 25 gpm is less than

those leak rates associated with the onset of unstable pipe

ruptures. This fact is also shown in FSAR figure 5.2-10. This value

of 25 gpm constitutes the design basis for the steam leak detection

system.

The mixing and liquid energy addition assumption changes in the

analysis do not affect this design basis. The analysis calculates

the resulting room temperature increase from a 25 gpm leak. In fact,

the new assumptions provide a more accurate yet conservative

prediction of room temperature increases. Therefore, operation of

the system is improved.

[[Page 64390]]

The proposed change leads to higher calculated room temperatures

to be used in the differential temperature setpoint calculations.

The engineering study was reviewed to determine if the higher

calculated temperatures would have a negative impact on the High

Energy Line Break and Leak Analysis environmental study which

provides the basis for equipment qualification.

In determining the room temperatures, the engineering study

considers ambient temperature setpoints at which the leaks will be

isolated. The proposed action will not change the ambient

temperature setpoints, and actuation of these instruments will

ensure that the results of the engineering study will not be

adversely affected. Therefore, no impact on equipment qualification

is being introduced by this change.

FSAR chapter 15 was reviewed for potential impacts on the

accident analyses. The 25 gpm leak outside containment is not

specifically analyzed in FSAR chapter 15. However, other conditions

which result in coolant leakage outside containment are analyzed in

section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System

Piping Break Outside Containment). As stated in the NRC SER, the 25

gpm RWCU leak rate is bounded by the analysis in FSAR section

15.6.4. FSAR section 15.6.2 also states that leak detection

actuations will initiate operator actions, a fact that is not

affected by the proposed change. Therefore, based on a review of

FSAR chapter 15 it was concluded that no impact on the analyzed

accident scenarios is created by the proposed change.

Based on the above discussions, it is demonstrated that the

proposed change will not adversely impact system function or

equipment. System performance will actually be improved since the

new setpoints eliminate spurious isolations resulting from a less

accurate model. The setpoint change has no impact on any equipment

important to safety or any accidents previously analyzed in the

FSAR. Therefore, the proposed change does not involve a significant

increase in the probability of occurrence or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed action does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Neither the system design basis nor the system function will be

adversely affected. System performance will be enhanced since

spurious differential temperature actuations will be reduced as a

result of using the more accurate, yet conservative, COTTAP model.

In addition to this, redundant temperature isolation function will

continue to be provided by the existing high ambient temperature

detectors.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed action does not involve a significant reduction in

a margin of safety. The Technical Specification basis for the

setpoints is to detect a leak below the flow rate corresponding to

critical crack size for the system piping. As stated previously, the

25 gpm flow rate is an acceptable flow rate and is used to calculate

the new temperatures.

Although the newly calculated RWCU penetration room temperatures

will be higher (due to the improved model), the isolation actuation

will be initiated by the high ambient temperature detectors before

the penetration room temperatures reach the newly calculated values,

as would happen under the old model. Therefore, system response is

not adversely affected.

The current temperature values lead to differential temperature

setpoints which are too low, causing spurious isolations. The use of

the new temperature values will reduce the number of spurious

isolations, reducing unnecessary challenges to safety systems during

normal plant operations.

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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: September 18, 1995

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Table 4.3.1.1-1, ``Reactor

Protection System Instrumentation Surveillance Requirements'' to

reflect the change in the calibration frequency for the Local Power

Range Monitor (LPRM) signal from every 1000 Effective Full Power Hours

(EFPH) to every 2000 Megawatt Days per Standard Ton (MWD/ST).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The change in the calibration frequency of the Local Power Range

Monitor (LPRM) signal does not make any physical change to the fuel

or the manner in which the fuel responds to a transient or accident.

The proposed TS change does not affect the fundamental method by

which the LPRMs are calibrated. Also, the LPRM calibration frequency

is not considered an initiator of any events analyzed in the SAR.

Therefore, calibrating the LPRMs on a different frequency will not

increase the probability of occurrence of an accident previously

evaluated in the SAR.

The resulting nodal power uncertainty does not exceed the nodal

power uncertainty accounted for in the existing Minimum Critical

Power Ratio (MCPR) Safety Limit; thus, the MCPR Safety Limit is not

affected by this TS Change, and, therefore, the initial conditions

of any accident are unchanged. Since the calibration frequency

change will not affect the course of any evaluated accident, the

consequences of an accident previously evaluated in the SAR will not

be increased.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The change in the calibration frequency of the Local Power Range

Monitor (LPRM) signal does not make any physical change to the plant

or the manner in which the equipment responds to a transient or

accident. The proposed TS change does not introduce a new mode of

plant operation and does not involve the installation of any new

equipment or instrumentation. The fuel will continue to be operated

to the same safety limits since the Minimum Critical Power Ratio

(MCPR) Safety Limit remains unchanged due to this TS change.

Therefore, the proposed TS change does not create the

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

accident previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The following TS Bases were reviewed for potential reduction in

the margin of safety:

2.0 Safety Limits and Limiting Safety System Settings;

3/4.1 Reactivity Control Systems;

3/4.2.1 Average Planar Linear Heat Generation Rate;

3/4.2.3 Minimum Critical Power Ratio:

3/4.2.4 Linear Heat Generation Rate;

3/4.3.1 Reactor Protection System Instrumentation;

3/4.3.6 Control Rod Block Instrumentation;

3/4.3.7.7 Traversing In-Core Probe System;

The GE Thermal Analysis Basis (GETAB) determination of the

Minimum Critical Power Ratio (MCPR) Safety Limit allows a maximum

total nodal uncertainty of the Traversing In-Core Probe (TIP)

readings of which the Local Power Range Monitor (LPRM).

Update uncertainty is a part. The change in LPRM calibration

frequency results in an LPRM Update uncertainty which, when combined

with the other uncertainties which comprise the total TIP readings

uncertainty, yields a total TIP readings nodal power uncertainty of

less than the allowed GETAB uncertainty. Thus the change in LPRM

[[Page 64391]]

calibration frequency will not affect the MCPR Safety Limit.

The LPRMs are utilized as input to the Average Power Range

Monitor (APRM) and Rod Block Monitor (RBM) systems. The primary

safety function of the APRM system is to initiate a scram during

core-wide neutron flux transients before the actual core-wide

neutron flux level exceeds the safety analysis design basis. This

prevents fuel damage from single operator errors or equipment

malfunctions. The APRMs are calibrated at least once per week to the

plant heat balance, utilize a radially and axially diverse group of

LPRMs as input and are utilized to detect changes in average, not

local, power changes. Therefore, the effects of changing the LPRM

calibration frequency on the APRM system responses will be minimal

due to any individual LPRM drift being practically canceled out (due

to diversity of input) and/or due to the frequent recalibration of

the APRMs to an independent power calculation (the heat balance).

Thus, changing the LPRM calibration frequency will not impact the

capability of the APRM system to perform the scram function, and

there is no impact on transient delta-CPRs.

The RBM system is utilized in the mitigation of a Rod Withdrawal

Error (RWE) event. The RBM system is designed to prevent the

operator from increasing the local power significantly when

withdrawing a control rod. Under Average Power Range Monitor - Rod

Block Monitor Technical Specifications/Maximum Extended Load Line

Limit Analysis (ARTS/MELLLA) on each selection of a control rod, the

average of the assigned, unbypassed LPRMs is adjusted to equal a

100% reference signal for each of the two RBM channels. Each RBM

channel automatically limits the local thermal margin changes by

limiting the allowable change in local average neutron flux to the

RBM setpoint. If the local average neutron flux change is greater

than that allowed by the RBM setpoint, within either RBM channel,

the rod withdrawal permissive is removed preventing further rod

movement. Since the change in local neutron flux is calculated from

the change in the average of the LPRM readings, and calibrated on

every rod selection to the reference signal, offsets in individual

LPRM readings due to calibration differences are effectively

eliminated for a given RBM setpoint. Therefore, the constraints on

the withdrawal of any given rod are unchanged, and there will not be

any increase in RWE delta-CPR.

Since the MCPR Safety Limit is unaffected and the delta-CPR

values are unchanged, the cycle CPR Operating limits are unchanged

due to this TS change. Therefore, the proposed change in the

frequency of LPRM signal calibration does not result in a reduction

in a margin of safety.

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

this review, it appears that the three standards of 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, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: May 3, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Surveillance Requirements

4.6.5.3.a and 4.6.5.4.a to modify specific requirements to perform

surveillance flow testing of the Standby Gas Treatment and Reactor

Enclosure Recirculation Systems from monthly to quarterly.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes do not involve any physical changes to

plant systems or equipment. The proposed TS changes only change the

Surveillance Requirements (SRs) surveillance test frequency

pertaining to flow testing of the SGTS and RERS from monthly to

quarterly. The periodic surveillance test frequencies provide

adequate assurance that the equipment tested will remain in an

operable condition. The test frequency interval for the flow testing

of the SGTS and RERS was determined from the regulatory position in

USNRC Regulatory Guide 1.52, ``Design, Testing, and Maintenance

Criteria for Post Accident Engineered-Safety-Feature Atmosphere

Clean-up System Air Filtration and Adsorption Units of Light-Water-

Cooled Nuclear Power Plants''. As stated in Regulatory Position

C.4.d, ''... each Engineered Safety Feature (ESF) atmosphere cleanup

train should be operated at least 10 hours per month, with the

heaters on (if so equipped), in order to reduce the buildup of

moisture on the absorbers and HEPA filters.''

System operation on a monthly basis for the purpose of

preventing moisture buildup on the absorbers as described in R.G.

1.52 is not required at Limerick due to the continuous dry

instrument air purge described previously in the Safety Assessment

section of this submittal. Therefore a change in the interval

between tests from monthly to quarterly will not result in moisture

accumulation which would reduce the capability of the absorber to

remove the iodine species from the exhaust air flow stream.

The SGTS components are common to both units and must be run

with the associated RERS for the surveillance test for each unit.

The currently specified test frequency results in the SGTS being run

at least twice per month or as many as eight (8) times per quarter

for this surveillance, in addition to other required system

surveillance tests which require the use of the components in this

system. A change in surveillance test frequency from monthly to

quarterly would reduce the wear on system components and thereby

reduce the associated system downtime for maintenance and repairs.

The consequent increased availability provides greater assurance

that the system will be able to perform its mitigation function

following any postulated accident.

Surveillance test frequency on a quarterly interval is

considered adequate to verify operability, as demonstrated by the

required quarterly test interval for other equipment important to

safety which have a similar function, such as the requirement for

quarterly verification of the isolation time of the secondary

containment and refueling area isolation valves, as required by LGS

TS Sections 4.6.5.2.1 and 4.6.5.2.2.

Therefore, the proposed TS changes do not involve an increase in

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 TS changes only involve changes to the frequency in

which the specified surveillances tests are performed. The proposed

TS changes do not physically change the design or intended function

of the systems, structures, or components associated with the SGTS

or RERS. There will be no change to the existing redundancy of

systems and components. The proposed change in surveillance test

frequency will not introduce the possibility of any failure

mechanisms of a different type than those already evaluated in the

SAR. The existing components will not be used in any different

manner and no new components will be added. Therefore with no

physical changes and no new or different manner of system operation,

no new failure mechanisms or equipment failure modes are created.

Therefore, the proposed TS changes do not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The margin of safety as defined in the LGS TS Bases has not been

reduced. The specific basis for the 31 day surveillance interval is

not given in the LGS TS Bases section nor in the LGS UFSAR Sections

6.5.1 or 9.4.2 which discuss the subject systems. However,

Regulatory Position C.4.d of Regulatory Guide 1.52, Revision 2,

relating to maintenance requirements, recommends:

Each ESF atmosphere cleanup train should be operated

at least 10 hours per month, with the heaters on (if so equipped),

in order to reduce the buildup of moisture on the absorbers and HEPA

filters.''

[[Page 64392]]

The Bases for Surveillance Requirements (SR) 3.6.4.3.1 in the

Standard Technical Specifications for General Electric Plants, BWR/

4, which corresponds to the subject LGS TS test, also notes the need

for ten (10) hours of operation per month for elimination of

moisture in the filters.

The basis for the requirement for a monthly test with the

heaters energized is clearly related to the desired elimination of

moisture in the filters and absorbers. However, LGS UFSAR Table 6.5-

2 states that LGS does not conform to R.G. 1.52, Position C.4.d

because the SGTS and RERS trains are ``continuously purged with dry

instrumentation air to prevent build-up of moisture.'' UFSAR

Sections 6.5.1.1.2 and 6.5.1.3.2 provide additional discussion of

this method of moisture control.

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, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: September 27, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would increase the Reactor Enclosure

Secondary Containment maximum inleakage rate. This change will also

impact secondary containment drawdown time and system flow rate

assumptions, thereby, affecting charcoal filter bed efficiency and post

accident dose analysis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

Changing the Reactor Enclosure post drawdown inleakage rate from

1250 cfm to 2500 cfm does not involve any changes to the function or

operation of any plant component or safety related system. The

Reactor Enclosure Recirculation System (RERS) and the Standby Gas

Treatment System (SGTS) will maintain their design function by

mitigating the radiological consequences of the analyzed accident

and mitigating the post LOCA temperatures within the Reactor

Enclosures. No analyzed accident initiating events are impacted, no

new accident initiators are created, and no new failure modes are

created. There are no changes to the redundancy, separation, quality

assurance or fire protection requirements for the associated

components and systems.

The proposed changes to the LGS adsorber bed residence time will

no longer fully meet the literal design guidance provided in

Regulatory Guide (RG) 1.52, ``Design, Testing, and Maintenance

Criteria for Post Accident Engineered-Safety-Feature Atmosphere

Cleanup System Air Filter and Adsorption Units of Light-Water-Cooled

Nuclear Power Plants,'' Revision 2, March 1978. This is because

LGS's unique, yet more conservative, adsorber bed design is not

addressed by the RG residence time design guidance. However, the LGS

SGTS charcoal adsorbers still conform to the design function

described in RG 1.52, based on the following: The LGS design with

increased inleakage will continue to conform to the three conditions

specified by RG 1.52, Position C.6.a, in order to maintain an

assigned decontamination efficiency of 99%; there is a conservative

amount of charcoal adsorber material provided by the LGS design,

based on calculations performed in accordance with RG 1.3

``Assumptions Used For Evaluating The Potential Radiological

Consequences of a Loss of Coolant Accident For Boiling Water

Reactors; and the LGS charcoal bed design is more conservative than

the RG 1.52 design guidance, based on data (i.e., Iodine Penetration

vs. Air Velocity) published by the charcoal manufacturer.

Therefore, the probability of occurrence and the consequences of

a malfunction of equipment important to safety is not increased.

Also, the probability of occurrence of an accident previously

evaluated is not increased. However, the proposed changes do affect

the leak tightness of the Unit 1 and Unit 2 Reactor Enclosure, which

increases the consequences of a postulated accident previously

evaluated.

Changing the Reactor Enclosure post drawdown inleakage rate from

1250 cfm to 2500 cfm will result in an increase in the calculated

LOCA/LOOP Design Basis Accident (DBA) off-site and on-site doses. 10

CFR Part 100, and 10 CFR Part 50 Appendix A, General Design Criteria

(GDC) 19, establish reference dose values used to determine site

suitability and provide reasonable assurance that the facility can

be operated following the analyzed accident without undue risk to

the health and safety of the public. The proposed TS changes will

increase the SGTS drawdown time from 2 minutes and 20 seconds to 15

minutes and 30 seconds. The drawdown time increase will not prevent

the RERS/SGTS from performing all of their safety related functions.

However, because it is conservatively assumed that all radioactive

material released during the drawdown period is unfiltered, and

because the drawdown period has been extended whereby more

unfiltered radioactive material is assumed to be released following

the DBA, there is a corresponding increase in the calculated

Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and

Control Room doses. It is also assumed that the SGTS exhausts at the

maximum inleakage rate throughout the entire DBA evaluation period

(i.e., 30 days) where an increase in the maximum inleakage rate

would also contribute to higher postulated EAB, LPZ, and Control

Room doses. However, the proposed calculated doses do not exceed 10

CFR Part 100, or 10 CFR Part 50, Appendix A, DGC 19 reference doses.

Since the proposed doses resulting from the changes remain below

10 CFR Part 100, and 10 CFR Part 50, Appendix A, these proposed

changes will not significantly increase the 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.

Changing the Reactor Enclosure post drawdown inleakage rate from

1250 cfm to 2500 cfm is not an accident initiator nor does it result

in the occurrence of an accident. The changes do not affect the

function or operation of any plant component or safety related

system nor do they create any new failure modes.

In addition, the proposed changes do not involve any changes to

the function or operation of any plant system or component nor will

they adversely affect the Reactor Enclosure post LOCA environmental

conditions. Furthermore, these changes will not create any new or

different failure modes for the equipment important to safety within

the Reactor Enclosure Secondary Containment.

Therefore, the possibility of an accident of a different type or

a different type of malfunction of equipment important to safety

than previously evaluated is not created.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

Changing the Reactor Enclosure post drawdown inleakage rate from

1250 cfm to 2500 cfm will result in reducing the margin of safety as

defined in the LGS Updated Final Safety Analysis Report (UFSAR)

relative to the off-site and on-site doses following a LOCA/LOOP

DBA, and an increase of the UFSAR specified system drawdown time.

From a system perspective, increasing the SGTS drawdown time from 2

minutes and 20 seconds to 15 minutes and 30 seconds will not prevent

the RERS/SGTS from performing all of their safety related functions.

There will be a postulated increase in the corresponding EAB, LPZ,

and Control Room doses, since it is assumed that fuel damage occurs

coincident with the LGS DBA (i.e, at time = 0), all radioactive

material released during the drawdown time is unfiltered, and the

drawdown time is proposed to be extended whereby more unfiltered

radioactive material could be released. It is also assumed that the

SGTS exhausts at the maximum inleakage rate throughout the entire

DBA evaluation period (i.e., 30 days) where an increase in the

maximum inleakage

[[Page 64393]]

rate would also contribute to higher postulated EAB, LPZ, and

Control Room doses. However, these calculated doses will remain

below 10 CFR Part 100, and 10 CFR Part 50, Appendix A, GDC 19

reference doses.

Therefore, these 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: October 1, 1996

Description of amendment request: The proposed amendment would

allow for a one-time extension of the surveillance intervals for the

containment isolation valve (CIV) seat leakage test, the isolation

valve seal water test, the boron injection tank leakage test, the

containment spray nozzle test, and the city water backup to the

auxiliary boiler feed pump test. These tests would be performed during

the refueling outage scheduled to begin in April 1997.

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:

Regarding the Containment Isolation Valve seat leakage and

Isolation Valve Seal Water tests:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The probability of a previously evaluated accident will

not increase because CIV leakage does not provide any role in

accident initiation. The CIVs provide containment isolation

following a design basis accident.

The consequences of an accident previously evaluated will not

significantly increase because the CIV leakage measurements contain

significant margin to a more restrictive criteria based on the

requested surveillance interval extension. As discussed in Section

II, ``Evaluation of Changes,'' [see application dated October 1,

1996] based on an evaluation of past CIV leak tests, the proposed

change will not result in an increase in containment leakage because

the measured leakage in previous CIV leak tests shows large margin

to a more restrictive criteria based on the requested surveillance

interval extension. Also, the latest test of IVSWS [isolation valve

seal water system] satisfied the established acceptance criteria.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed license amendment does not create the possibility

of a new or different kind of accident from any previously

evaluated. The proposed change only provides for a relatively short,

one-time extension of the current leak-test interval for certain

containment isolation valves. The proposed change does not involve

the addition of any new or different type of equipment, nor does it

involve operating equipment required for safe operation of the

facility in a manner different from that addressed in the Final

Safety Analysis Report. Therefore, the proposed change will not

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

any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed amendment does not involve a significant reduction

in a margin of safety. The proposed change, for a one-time extension

of the test interval, will not result in a significant reduction in

a margin of safety because the test interval is being extended by

only a short period and the measured leakage in previous CIV leak

tests shows large margin to a more restrictive criteria based on the

surveillance interval extension. In addition, the online leakage

monitoring capability of the WCCPPS [weld channel containment

penetration pressurization system] helps ensure that changes in CIV

leakage during the extension period will be detected. Therefore,

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

safety.

Regarding the Boron Injection Tank (BIT) leakage test:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The proposed change will not degrade the

integrity of the BIT piping outside containment because no time

dependent failure trends were observed in the review of past test

results. The probability of a previously evaluated accident will not

be increased because BIT leakage does not provide any role in

accident prevention. The BIT leakage test only verifies that the BIT

and associated piping meet specified leakage limits.

The consequences of an accident previously evaluated will not

significantly increase because the BIT leakage test results show

large margins to the allowable leakage limit.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed license amendment does not create the possibility

of a new or different kind of accident from any previously

evaluated. The proposed change does no[t] involve the addition of

any new or different type of equipment, nor does it involve

operating equipment required for safe operation of the facility in a

manner that's different from that addressed in the Final Safety

Analysis Report. Also, the increased surveillance interval (one-time

only) will not adversely affect the integrity of the BIT piping and

will not result in any new failure modes. Therefore, the proposed

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

accident from any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed license amendment does not involve a significant

reduction in a margin of safety. Because of the large margin between

the previous test and the allowable leak rate limit, the proposed

change, for a one-time extension of the test interval, for the BIT

leakage test does not adversely affect the performance of any safety

related system, component, and does not result in increased severity

of any of the accidents considered in the Final Safety Analysis

Report. Based on past test results, the one-time extension of the

leak test interval does not involve a significant reduction in a

margin of safety.

Regarding the Containment Spray Nozzle test:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. As discussed in Section II, ``Evaluation of

Changes,'' [see application dated October 1, 1996] based on an

evaluation of past test results the proposed change will not degrade

the reliability of the Containment Spray Nozzles because no time

dependent failure trends were observed in the data review. The

probability of a previously evaluated accident will not be increased

because the Containment Spray Nozzles do not provide any role in

accident prevention. The Containment Spray Nozzles provide a uniform

spray distribution for containment cooling following postulated

post-accident conditions.

The consequences of an accident previously evaluated will not

increase because the Containment Spray Nozzle reliability is not

degraded by this change.

[[Page 64394]]

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed license amendment does not create the possibility

of a new or different kind of accident from any previously

evaluated. The proposed change does not involve the addition of any

new or different type of equipment, nor does it involve operating

equipment required for safe operation of the facility in a manner

that is different from that addressed in the Final Safety Analysis

Report. Also, the increased surveillance interval (one-time only)

w[i]ll not adversely affect the functioning of the Containment Spray

Nozzles and will not result in any new failure modes. Therefore, the

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed license amendment does not involve a significant

reduction in a margin of safety. The proposed change, for a one-time

extension of the test interval, for the Containment Spray Nozzles

does not adversely affect the performance of any safety related

system, component, or instrument, or safety system setpoints and

does not result in increased severity of any of the accidents

considered in the Final Safety Analysis Report. Based on past test

results, the one-time extension of the functional test interval will

not adversely affect the functioning of the Containment Spray

Nozzles. Therefore, this change does not create a significant

reduction in a margin of safety.

Regarding the City Water Backup test:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The proposed change will not degrade the

reliability of the City Water Backup Supply Valves for the AFW

[auxiliary feedwater] System because no time dependent failure

trends were observed in the review of past test results. The

probability of a previously evaluated accident will not increase

because the City Water Backup Supply Valves for the AFW System do

not provide any role in accident prevention. The City Water Backup

Supply Valves for the AFW System only provides a diverse source of

water for the AFW system.

The consequences of an accident previously evaluated will not

significantly increase because the City Water Backup Supply Valves

for the AFW System are not assumed to function to mitigate any

analyzed accident.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed license amendment does not create the possibility

of a new or different kind of accident from any previously

evaluated. The proposed change does not involve the addition of any

new or different type of equipment, nor does it involve operating

equipment required for safe operation of the facility in a manner

that is different from that addressed in the Final Safety Analysis

Report. Also, the increased surveillance interval (one-time only)

will not adversely affect the functioning of the City Water Backup

Supply Valves for the ABFP [auxiliary boiler feedpump] and will not

result in any new failure modes. Therefore, the proposed change will

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

from any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed amendment does not involve a significant reduction

in a margin of safety. The proposed change, for a one-time extension

of the test interval, for the City Water Backup Supply Valves for

the ABFP does not adversely affect the performance of any safety

related system, component, or instrument, or safety system setpoints

and does not result in increased severity of any of the accidents

considered in the Final Safety Analysis Report. Based on past test

results, the one-time extension of the functional test interval will

not adversely affect the functioning of the City Water Backup Supply

Valves for the AFW System. Therefore, this change does not create a

significant reduction in a margin of safety.

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

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

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

amendment 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: S. Singh Bajwa, Acting

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: October 25, 1996

Description of amendment request: The proposed change to Hope Creek

Technical Specification (TS) 3/4.1.3.5, ``Control Rod Scram

Accumulator'', would: 1) permit a separate entry into a Technical

Specification action statement for each inoperable control rod; 2)

provide more specific applicability for required actions in operational

condition 1 or 2 with one inoperable control rod scram accumulator

(reactor pressure of greater than or equal to 900 psig would be

specified); 3) provide more specific actions (verify charging water

pressure) for two or more inoperable control rod scram accumulators and

reactor pressure is greater than or equal to 900 psig; 4) provide more

specific actions when reactor pressure is less than 900 psig and one or

more control rod scram accumulators are inoperable (verify insertion of

control rods associated with inoperable accumulators and verify that

charging water header pressure is greater than or equal to 940 psig);

and 5) provide specific actions in operational condition 5 with one or

more withdrawn control rods inoperable; and 6) eliminate the

requirements to perform a 18-month channel functional test of the leak

detectors and the 18-month channel calibration of the pressure

detectors.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The change incorporates the appropriate content of the improved

BWR/4 Standard Technical Specifications, NUREG-1433, for Control Rod

Scram Accumulators.

The proposed Technical Specification and required Action

completion times are consistent with or more conservative than those

approved for use in the improved Technical Specifications for

inoperable control rod scram accumulators. In addition, the proposed

surveillance requirements for the control rod scram accumulators are

sufficient to adequately demonstrate operability as stated in the

Bases for the improved Technical Specifications. Further, the

proposed changes enhance the current Hope Creek Technical

Specifications by reflecting improved techniques collectively

learned by the industry. Therefore, the proposed changes do not

significantly increase the risk or consequences of any accidents

previously evaluated.

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

or different kind of accident from any previously evaluated.

Neither the mechanism for initiating or completing a scram is

modified by this proposed change. There are no physical changes to

plant equipment proposed in the application. The proposed change

does not create a means by which the scram function could be impeded

or prevented. The proposed change is functionally equivalent to the

current Technical Specifications, but provides additional

operational flexibility to diagnose and resolve equipment issues

that do not impact operability of the control rods before taking

proscriptive actions which

[[Page 64395]]

result in significant plant transients (i.e. full power scram).

3. The proposed change does not involve a significant reduction

in a margin of safety.

The operability of the accumulators and the scram function of

the control rod drive system protects the Safety Limit Minimum

Critical Power Ratio as well as the 1% cladding plastic strain fuel

design limit. The proposed change does not reduce a margin of safety

as defined in the Bases of the Technical Specification since the

proposed change does not affect the maximum allowable scram times

for control rods, nor does it change the maximum allowable number or

minimum separation of inoperable control rods. The proposed change

does not modify any instrument setpoints or functions. The proposed

change will either maintain the present margins of safety or

increase them, by reducing the need for unnecessary challenges to

the reactor protection system and resulting plant shutdowns, while

still maintaining the capability to complete a reactor scram.

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

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: October 29, 1996

Description of amendment request: The proposed amendment would

revise the mode of applicability for the motor-driven auxiliary

feedwater (AFW) pump actuation on opening of the main feedwater (MFW)

pump breakers to correct an error introduced during Amendment No. 61.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The less restrictive changes discussed in Section C.1 [of the

licensee's application] do not involve a significant hazards

consideration as discussed below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated. The proposed

changes only correct an error which was introduced in Amendment No.

61 to the Ginna Station technical specifications. The changes revert

the mode of applicability for the motor-driven AFW pump actuation on

the opening of the MFW pump breakers to what existed previously. The

change is essentially correction of a typographical error that was

caused through use of the electronic version of NUREG-1431 in

preparation of the Ginna Station ITS [Improved Technical

Specifications]. There have been no subsequent plant modifications

or changes to the accident analysis which would invalidate the

previous NRC acceptance of only requiring this Function above 5%

power. The accident analyses do not credit automatic initiation of

AFW on MFW pump trip in MODE 2. As such, these changes do not impact

initiators or analyzed events or assumed mitigation of accident or

transient events. Therefore, these changes do not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

2. Operation of Ginna Station in accordance with the proposed

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

accident from any accident previously evaluated. The proposed changes

do not involve a physical alteration of the plant (i.e., no new or

different type of equipment will be installed) or changes in the

methods governing normal plant operation which existed prior to

Amendment No. 61. The proposed changes will not impose any new or

different requirements. Thus, this change does not create the

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

previously evaluated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. The proposed changes will not reduce a margin of plant

safety because there have been no subsequent plant modifications or

changes to the accident analysis which would invalidate the previous

NRC acceptance of only requiring this Function above 5% power. As

such, no question of safety is involved, and the change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005

NRC Project Director: S. Singh Bajwa, Acting

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: October 29, 1996

Description of amendment request: The proposed amendment would

revise the Required Actions for the auxiliary feedwater (AFW) pump

actuation on Steam Generator Level (SG) - Low Low logic to be

consistent with those specified in NUREG-1431.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The less restrictive changes discussed in Section C.1 [of the

licensee's application] do not involve a significant hazards

consideration as discussed below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated. The proposed

changes with respect to the Required Actions for AFW actuation on SG

Level - Low Low logic provide consistency with NUREG-1431 by

requiring an inoperable channel to be placed in the tripped

condition within 6 hours. The affected logic then requires 1 of 2

channels in order to actuate such that there is no impact on any

initiators or analyzed events or assumed mitigation of accident or

transient events. Therefore, these changes do not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

2. Operation of Ginna Station in accordance with the proposed

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

of accident from any accident previously evaluated. The proposed

changes do not involve a physical alteration of the plant (i.e., no

new or different type of equipment will be installed) or changes in

the methods governing normal plant operation. The proposed changes

will not impose any new or different requirements. Thus, this change

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

accident from any accident previously evaluated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. The proposed changes will not reduce a margin of plant

safety because the AFW actuation on SG Level - Low Low still remains

capable of performing its function with an inoperable channel placed

in the tripped configuration. These changes are also consistent with

those provided in NUREG-1431. As such, no question of safety is

involved, and the change does not involve a significant reduction in

a margin of safety.

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

this review, it appears that the three

[[Page 64396]]

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: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005

NRC Project Director: S. Singh Bajwa, Acting

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

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

Virginia

Date of amendment request: September 4, 1996

Description of amendment request: The proposed amendment to the

Technical Specifications would allow the use of four lead test

assemblies (advanced zirconium-based alloys) in the North Anna, Units 1

and 2, reactor cores.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the four FCF [Framatome Cogema Fuels] lead test

assemblies will not:

1.Involve a significant increase in the probability of

occurrence or the consequences of an accident previously evaluated.

The FCF lead test assemblies are very similar in design to the

Westinghouse fuel that comprises the remainder of the core. The

reload core design for North Anna cycles which incorporate the lead

test assemblies will meet all applicable design criteria. In

addition, the performance of the ECCS [emergency core cooling

system] at North Anna Units 1 and 2 will not be affected by the

insertion of the four lead test assemblies, so the criteria of 10

CFR 50.46 will be satisfied for use of these assemblies with fuel

rods, guide thimble tubes, and instrumentation tubes fabricated with

advanced zirconium-based alloys. The use of these fuel assemblies

will not result in a change to the North Anna Units 1 and 2 reload

design and safety analysis limits. The existing safety analyses

based on the resident Westinghouse fuel will remain applicable for

cycles which incorporate the lead test assemblies. Therefore,

neither the probability of occurrence nor the consequences of any

accident previously evaluated is significantly increased.

2. Create the possibility for a new or different type of

accident from any accident previously evaluated. The FCF lead test

assemblies are very similar in design (both mechanical and

composition of materials) to the resident Westinghouse fuel. North

Anna cores which incorporate the lead test assemblies will be

designed to meet all applicable design criteria and ensure that all

pertinent licensing basis criteria are met. Demonstrated adherence

to these standards and criteria precludes new challenges to

components and systems that could introduce a new type of accident.

North Anna safety analyses based on the resident Westinghouse fuel

will remain applicable for cores containing the lead test

assemblies. All design and performance criteria will continue to be

met and no single failure mechanisms have been created. In addition,

the use of these fuel assemblies does not involve any alteration to

plant equipment or procedures which would introduce any new or

unique operational modes or accident precursors. Therefore, the

possibility for a new or different kind of accident from any

accident previously evaluated is not created.

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

use of the FCF lead test assemblies does not change the performance

requirements on any system or component such that any design

criteria will be exceeded, and will not cause the core to operate in

excess of pertinent design basis operating limits. North Anna reload

core designs for cycles which incorporate the lead test assemblies

will specifically evaluate any pertinent differences between the

lead test assemblies and the resident fuel, and will take into

consideration the normal core operating conditions allowed in the

Technical Specifications. Safety analyses based on the resident

Westinghouse fuel will remain applicable for cores incorporating the

FCF lead test assemblies. Analyses or evaluations will be performed

each cycle to confirm that the criteria in 10 CFR 50.46 will be met.

Therefore, the margin of safety as defined in the Bases to the North

Anna Units 1 and 2 Technical Specifications is not significantly

reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Acting Project Director: Mark Reinhart

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

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

Virginia

Date of amendment request: November 6, 1996

Description of amendment request: The proposed changes will modify

the requirements for isolated loop startup to permit filling of a

drained isolated loop via backfill from the reactor coolant system

through partially open stop loop valves.

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:

Specifically, operation of the North Anna Power Station [in]

accordance with the proposed changes will not:

1. Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated. The

probability of occurrence of a positive reactivity addition accident

is not being increased by the proposed Technical Specification

change. The proposed restrictions on boron concentration and mixing,

reactor coolant system inventory and reactivity and count rate

monitoring provide a level of protection against reactivity addition

accidents which is equivalent to that currently in place.

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

from any accident previously evaluated. The proposed change does not

introduce any new or unique failure modes or accident precursors.

Eliminating the operability requirements for the loop stop valve

interlocks does not create any new or different kind of accident

scenario. Loop startup accidents in the various modes of operation

have been analyzed. Operation of the loop stop valves will not

change. New requirements have been imposed for the case of

backfilling a drained loop from the reactor coolant system to ensure

that core cooling and reactivity control are preserved throughout

the backfill evolution.

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

new Technical Specification loop isolation and startup requirements

for temperature, boron concentration, and shutdown margin fulfill

the function of the loop stop valve interlocks. Therefore, the

margin of safety as defined in any Technical Specification bases is

not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Mark Reinhart (Acting)

[[Page 64397]]

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: October 31, 1996

Description of amendment request: The proposed amendment would

revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications

(TS) by deleting the requirement for an annual submittal of a

description of changes made pursuant to 10 CFR 50.59. Consistent with

10 CFR 50.59(b)(2), a description of changes will subsequently be

included with the KNPP Updated Safety Analysis Report (USAR) update in

accordance with 10 CFR 50.71(e). Additionally, the proposed amendment

would correct minor administrative inconsistencies in the TS Table of

Contents and in a footnote reference.

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. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c)The NRC staff's review is

presented below:

On August 31, 1992 (57 FR 39353), the NRC amended 10 CFR

50.59(b)(2) to reduce the regulatory burden on nuclear licensees. This

action revised the requirements for the annual submission of reports

for facility changes under 10 CFR 50.59. This action did not affect the

substance of the evaluation or the documentation required for 10 CFR

50.59 type changes. It only affected the interval for submission of the

information to the NRC. Instead of submitting the information annually,

the information can be submitted on a refueling cycle basis, provided

the interval between successive reports does not exceed 24 months.

In order to take advantage of this reduction in regulatory burden,

the licensee has proposed an amendment to remove the submittal of a

report of facility changes under 10 CFR 50.59 from the Technical

Specification list of annual reporting requirements. Additionally, the

licensee has proposed corrections to minor administrative

inconsistencies in the TS Table of Contents and in a footnote

reference. The proposed changes are administrative only and do 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.

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: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497

NRC Project Director: Gail H. Marcus

NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station,Plymouth County, Massachusetts

Date of application for amendment: May 1, 1996, as supplemented

August 12, 1996.

Brief description of amendment: The amendment approves relocation

of the administrative controls related to the quality assurance review

and audit requirements of Section 6, Technical Specifications 6.5.B.8,

``Nuclear Safety Review and Audit Committee-Audits,'' from the Pilgrim

Station Technical Specifications to the Boston Edison Quality Assurance

Manual (BEQAM). This change is in accordance with the guidance

contained in NRC Administrative Letter 95-06, ``Relocation of Technical

Specification Administrative Controls Related to Quality Assurance.''

In addition, the Safety Evaluation includes the NRC staff review and

approval of the BEQAM changes in support of this amendment.

Date of issuance: November 12, 1996

Effective date: November 12, 1996

Amendment No.: 168

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28605) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 12, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of application for amendments: August 29, 1996, as

supplemented on September 20, 1996, and October 4, 1996.

Brief description of amendments: The amendments change the

Technical Specifications to implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory Guide 1.163, ``Performance-Based

Containment Leakage-Test Program,'' with an exception as detailed in

the licensee's application.

Date of issuance: November 12, 1996

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 175 and 162

[[Page 64398]]

Facility Operating License Nos. DPR-39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52964). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 12, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of application for amendment: August 14, 1996, as supplemented

October 18, 1996, and related application of January 18, 1996

Brief description of amendment: The amendment revises the technical

specifications (TS) to allow one-cycle deferral of the inspection of

reactor coolant pump (RCP) flywheels.

Date of issuance: November 7, 1996

Effective date: November 7, 1996

Amendment No.: 175

Facility Operating License No. DPR-20. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 24, 1996 (61

FR 50054) The October 18, 1996, letter provided an updated TS page.

This change was within the scope of the original application and did

not change the staff's initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated November 7, 1996.No

significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423

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: December 14, 1994, as

supplemented by letters dated May 16 and August 29, 1996

Brief description of amendments: The amendments will incorporate

guidance and recommendations for diesel generators contained in NUREG-

1366, ``Improvements to Technical Specifications Surveillance

Requirements,'' Generic Letter (GL) 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operations,'' GL 94-01, ``Removal of Accelerated

Testing and Reporting Requirements for Emergency Diesel Generators,''

and NUREG-1431, ``Revised Standard Technical Specifications for

Westinghouse PWRs.''

Date of issuance: November 12, 1996

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

30 days

Amendment Nos.: 170 and 152

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28612) The August 29, 1996, letter provided clarifying information that

did not change the scope of the December 14, 1996, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 12, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station,

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: August 1, 1996

Brief description of amendment: The amendment revises the technical

specifications to incorporate requirements for limiting the time that

the hydrogen mixing isolation valves on the drywell are open. The

amendment also changes the time from 7 days to 31 days to determine the

cumulative time the valves are open.

Date of issuance: November 12, 1996

Effective date: November 12, 1996

Amendment No.: 89

Facility Operating License No. NPF-47. The amendment revised the

Technical Specifications/operating license.

Date of initial notice in Federal Register: September 25, 1996 (61

FR 50343) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 12, 1996.No significant

hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Entergy Operations, Inc., System Energy Resources, Inc.,

SouthMississippi Electric Power Association, and Entergy

Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station,

Unit 1, Claiborne County, Mississippi

Date of application for amendment: May 9, 1996, as supplemented by

letter dated August 27, 1996.

Brief description of amendment: The amendment changed Surveillance

Requirements (SRs) 3.4.4.3, Safety/Relief Valves, 3.5.1.7, Automatic

Depressurization System Valves, and 3.6.1.6.1, Low-Low Set Valves, of

the Technical Specifications and allows the licensee to perform the

surveillance of the relief mode of operation of the safety/relief

valves on the main steam lines without physically lifting the disk of a

valve off the seat at power. The changes stated that the required

operation of the valve to verify is that the relief-mode actuator

strokes when the valve is manually actuated and the frequency of the

surveillances are in accordance with the inservice testing program for

the valves.

Date of issuance: November 18, 1996

Effective date: November 18, 1996

Amendment No: 130

Facility Operating License No. NPF-29. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47971) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 18, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

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

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

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

Connecticut

Date of application for amendment: June 3, 1996, as supplemented

October 23, 1996

Brief description of amendment: The amendment clarifies a

restriction on shutdown margin monitor operability while changing

operational modes, so that it only limits reactivity changes caused by

boron dilution and rod withdrawal. The amendment also corrects a

technical specification numerical reference so that the specification

number cited is in agreement with Amendment 99, dated December 29,

1994.

Date of issuance: November 14, 1996

[[Page 64399]]

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

within 60 days.

Amendment No.: 131

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

Technical Specifications.

Date of initial notice in Federal Register: June 20, 1996 (61 FR

31559) The October 23, 1996, letter provided clarifying information

that did not change the scope of the June 3, 1996, application and the

initial proposed no significant hazards consideration determination.The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated November 14, 1996No significant hazards

consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385.

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: May 30, 1996

Brief description of amendment: The proposed change to the

anticipated transient without scram recirculation pump trip logic for

the James A. Fitzpatrick Nuclear Power Plant allows for a high pressure

trip setpoint which is dependent upon the number of safety/relief

valves which are out of service.

Date of issuance: November 7, 1996

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

30 days.

Amendment No.: 237

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34896) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 7, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: May 30, 1996, as supplemented

October 17, and November 8, 1996

Brief description of amendment: The proposed amendment changes the

FitzPatrick safety limit minimum critical power ratio from its current

value of 1.07 for two recirculation loop operation to 1.09 and from

1.08 to 1.10 for single recirculation loop operation for the Cycle 13

operation.

Date of issuance: November 14, 1996

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

30 days.

Amendment No.: 238

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34896) The October 17 and November 8, 1996 letters provided

supplemental information that did not change the initial no significant

hazards consideration determination.The Commission's related evaluation

of the amendment is contained in a Safety Evaluation dated November 14,

1996.No significant hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph

M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: August 23, 1996, as supplemented by

letters dated September 16, November 6, 11 and 14, 1996

Brief description of amendment: The amendment changes the Technical

Specifications (TS) to allow installation of laser welded elevated

tubesheet sleeves. Specifically, the amendment is for one cycle only

for Farley Unit 2. Permanent, generic TS changes for Westinghouse laser

welded sleeves for both units will be submitted prior to the next Unit

1 refueling outage currently scheduled for spring 1997.

Date of issuance: November 20, 1996

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

30 days

Amendment No.: 117

Facility Operating License No. NPF-8: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47982) The September 16, November 6, 11 and 14, 1996, letters

provided clarifying information that did not change the scope of the

August 23, 1996, application and the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 20, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Southern California Edison Company, et al., Docket Nos. 50-361 and

50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,

San Diego County, California

Date of application for amendments: July 17, 1995.

Brief description of amendments: These amendments revise the

frequency of surveillance requirements for certain plant protective

system instrumentation contained in Technical Specifications (TS)

3.3.1, ``Reactor Protective System (RPS) Instrumentation - Operating,''

TS 3.3.2, ``Reactor Protective System (RPS) Instrumentation -

Shutdown,'' TS 3.3.3, ``Control Element Assembly Calculators (CEACs),''

TS 3.3.4, ``Reactor Protective System (RPS) Logic and Trip

Initiation,'' TS 3.3.5, ``Engineered Safety Features Actuation System

(ESFAS) Instrumentation,'' and TS 3.3.6, ``Engineered Safety Features

Actuation System (ESFAS) Logic and Manual Trip.''

Date of issuance: November 18, 1996

Effective date: November 18, 1996, to be implemented within 30 days

of the date of issuance.

Amendment Nos.: Unit 2 - 133 ; Unit 3 - 122

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45185) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 18, 1996.No significant

hazards consideration comments received: No.Temporary

Local Public Document Room location: Science Library, University

of California, P. O. Box 19557, Irvine, California 92713

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of application for amendment: September 4, 1996

Brief description of amendment: This amendment revises Technical

[[Page 64400]]

Specification (TS) 6.2.3, ``Facility Staff Overtime,'' by removing

specific overtime limits and working hours and by adding procedural

controls to perform a monthly review of overtime hours.

Date of issuance: November 8, 1996

Effective date: November 8, 1996, to be implemented not later than

90 days after issuance

Amendment No.: 212

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

Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52970) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 8, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

Callaway County, Missouri

Date of application for amendment: July 18, 1996

Brief description of amendment: The amendment adopts ASTM D-3803-

1989 as the laboratory testing standard for charcoal samples from the

charcoal adsorbers in the auxiliary/fuel building emergency exhaust

system.

Date of issuance: November 13, 1996

Effective date: November 13, 1996, to be implemented within 30 days

of the date of issuance.

Amendment No.: 118

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 14, 1996 (61 FR

42285) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 13, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Dated at Rockville, Maryland, this 26th day of November 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II Office of Nuclear Reactor

Regulation

[Doc. 96-30714 Filed 12-3-96; 8:45 am]

BILLING CODE 7590-01-F