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 December 20, 1996, through January 3, 1997.

The last biweekly notice was published on January 2, 1997 (62 FR 121).

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 February 14, 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 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.

[[Page 2186]]

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.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of amendment request: November 4, 1996, as supplemented on

December 4, 1996.

Description of amendment request: The proposed amendment would

permit Byron, Unit 1, and Braidwood, Unit 1, to remove sheathing filler

grease in the tendon sheathing for up to 35 tendons in advance of the

steam generator replacement outages.

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 prestressing tendons are passive components that form part

of the containment structure. As passive components, there are no

tendon failure modes that could act as accident initiators or

precursors.

Consequently, the proposed change to remove a portion of the

tendon sheathing filler grease will not increase the probability of

an accident previously evaluated.

The tendons, in their passive role, function to limit the

consequences of accidents previously evaluated, and their continued

integrity is important to the ability of the containment to mitigate

design basis accidents. Structural degradation of the containment is

a predictable process that can be monitored by a comprehensive

containment tendon monitoring program as required by Technical

Specification Surveillance Requirement 4.6.1.6. The monitoring

program is based on proposed Revision 3 of Regulatory Guide 1.35,

``Inservice Surveillance of Ungrouted Tendons in Prestressed

Concrete Containment Structures,'' April 1979.

The tendon surveillances conducted at both Byron and Braidwood

have consistently shown that structural integrity of the tendon

system has been maintained, including adequate corrosion protection

for the tendon wires and end anchorage components, and there has

been no evidence of grease leakage from the tendon sheathings. While

a number of below-grade hoop tendons have shown signs of water

intrusion, the tendons that will have grease removed are above-grade

and are not expected to experience water intrusion.

A review of domestic nuclear facility experience found cases

where large grease voids existed for periods longer than requested

under the proposed change without resultant corrosion in those

tendon systems. A case where tendon wires removed from a

decommissioned plant were exposed to an environment more severe than

expected in a sealed tendon sheath did not show signs of corrosion.

These experiences demonstrate the effectiveness of the initial

corrosion protection systems applied to the tendons and the

effectiveness of partial grease protection in the tendon sheathing.

Based on the above cases, it can be concluded that the removal

of the filler grease (grease voids greater than 5 percent) from the

tendon sheathing in up to thirty-five tendons for a limited period

will not adversely affect the integrity of the tendons or the

capability of the tendon system to fulfill its design basis

function.

The removal process will only remove the grease not directly

adhering to the tendons. The grease remaining will be adequate to

protect the tendons during the relatively short period of partial

grease removal. Therefore, no changes in the tendon properties would

be expected, and the consequences of design basis accidents

previously evaluated will not be affected by the proposed change.

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

or different kind of accident from any accident previously

evaluated.

[[Page 2187]]

The proposed change only affects the tendon sheathing filler

grease void limits of TSSR 4.6.1.6. No new equipment is being

installed and no existing equipment is being modified. Operation

with a grease void in excess of current requirements does not alter

system configurations such that any new or different accidents can

be initiated. Therefore, no new or different accident initiators or

precursors are being introduced, and the proposed change will not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The margin of safety applicable to the proposed change is

defined by the difference between the design pressure of the

containment and the point at which the containment would actually

fail. The design pressure of the containment is 50 psi. As a result

of conservatism inherent in the design techniques and in the

material selections made for the Byron and Braidwood containments, a

substantial margin to failure exists in the containment. This margin

is discussed in Subsection 3.8.1.8 of the Updated Final Safety

Analysis Report. It is noted therein that the ultimate capacity of

the concrete shell is 125 psi, corresponding to the initiation of

yield in the hoop post-tensioning tendons in conjunction with

yielding of the reinforcement near the mid-height of the containment

wall.

It is also noted in Subsection 3.8.1.8 that the ultimate

capacity of a containment electrical penetration is 108 psi. While

this value is substantially greater than the 50 psi required of the

design, it is lower than the 125 psi at which failure of the

containment wall section would be predicted. Therefore, tendon

strength is not the limiting factor in the margin of safety inherent

in the containment.

As previously discussed, no degradation of the tendons is

expected to occur as a result of the proposed TS change. Further,

the tendon strength is not the limiting factor in the containment

ultimate capacity, which is substantially greater than the

requirement placed on the containment design by the plant design

basis. Therefore, the proposed change will not reduce the margin of

safety designed into Byron and Braidwood.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington Illinois 60481.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendment request: December 6, 1996

Description of amendment request: The proposed amendment would

allow a single control rod to be moved when the plant is in HOT

SHUTDOWN and COLD SHUTDOWN condition provided the one-rod-out interlock

is OPERABLE and the reactor mode switch is in the refuel position.

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) Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

This revision would allow a single control rod to be withdrawn

under control of the reactor mode switch position one-rod-out

interlock in OPERATIONAL MODES 3 or 4. This interlock is explicitly

assumed in the safety analysis for control rod removal error during

refueling. A prompt reactivity excursion could potentially result in

fuel failure. The one-rod-out interlock, together with the

requirements for adequate SHUTDOWN MARGIN (SDM), provides protection

against prompt reactivity excursions by preventing withdrawal of

more than one control rod and ensuring the core remains subcritical

with any one control rod withdrawn. The addition of surveillance

requirements for the one-rod-out interlock will assure the interlock

is OPERABLE prior to withdrawal of a control rod in OPERATIONAL

MODES 3 and 4. Although this change will increase the frequency of

single control rod withdrawals in OPERATIONAL MODES 3 and 4, the

probability of previously analyzed accidents, including control rod

withdrawal error, is not affected because the same actions are

required, although they are now conducted in different OPERATIONAL

MODES.

The consequences of previously analyzed accidents in OPERATIONAL

MODES 3 and 4 are not affected by this proposed change. The SDM

requirements of TS 3.3.A assure the reactor is maintained

subcritical when all control rods are fully inserted, without

crediting the single control rod having the highest reactivity worth

which is assumed to be fully withdrawn. The one-rod-out interlock of

the reactor mode switch Refuel position permits only a single

control rod to be withdrawn. The proposed change will not affect the

potential for attaining criticality in OPERATIONAL MODES 3 and 4 or

effect the initial conditions assumed in any design basis accident

analysis.

Based on this, the probability or consequences of any accident

previously evaluated is not increased by the proposed changes.

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

from any accident previously evaluated because:

Single control rods can be withdrawn to permit control rod

recoupling in OPERATIONAL MODES 3 and 4 under existing TS. The

proposed change will merely expand this allowance to other control

rod maintenance and testing activities performed in OPERATIONAL

MODES 3 and 4. The revision to Specification 3/4.10.A provides

additional assurance that the one-rod-out interlock is OPERABLE in

OPERATIONAL MODES 3 and 4.

The additional control rod maintenance and testing activities

which could be performed in OPERATIONAL MODES 3 and 4 are permitted

by the existing TS in OPERATIONAL MODES 1, 2 and 5. Examples of

activities which could be performed include venting of control rods

following a reactor scram or control rod drive system outage, normal

control rod insertion/withdrawal timing and adjustment, control rod

scram time testing and control rod friction testing.

Based on this, the proposed changes do not create the

possibility of a new or different kind of accident from those

previously evaluated.

Specification 3/4.10.A is revised to ensure the one-rod-out

interlock is OPERABLE, enhancing the assurance that the plant will

prevent the withdrawal of more than one control rod in the manner

currently assumed. Expanding the applicability of this existing

requirement to OPERATIONAL MODES 3 and 4 similarly does not create

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

previously evaluated.

3) Involve a significant reduction in the margin of safety

because:

The TS currently permit single control rod withdrawal for the

purpose of control rod recoupling when in OPERATIONAL MODES 3 or 4

if the one-rod-out interlock is OPERABLE. This change merely allows

additional activities for which a single control rod may be

withdrawn in OPERATIONAL MODES 3 or 4, with the same restriction

that the one-rod-out interlock is OPERABLE.

While the TS currently allow limited control rod withdrawal in

OPERATIONAL MODES 3 and 4 provided the one-rod-out interlock is

OPERABLE, no explicit surveillance requirements for the one-rod-out

interlock exist while in OPERATIONAL MODES 3 or 4. The proposed

changes to the Applicability statement in TS 3/4.10.A will result in

applicability of the Surveillance Requirements for the one-rod-out

interlock whenever control rod withdrawal is performed in

OPERATIONAL MODES 3 and 4.

Together, the OPERABILITY requirements for the one-rod-out

interlock and the SDM requirements of TS 3.3.A will continue to

ensure that the reactor will be maintained subcritical during single

control rod withdrawals. Therefore, this change will not involve a

significant reduction in the margin of safety.

[[Page 2188]]

As described, the proposed amendment for Dresden and Quad Cities

Stations will not reduce the availability of systems required to

mitigate accident conditions. Neither are new or significantly

different modes of operation proposed. Therefore, the proposed

changes do not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Area

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

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

Illinois 61021

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

Date of amendment request: October 2, 1996

Description of amendment request: Relocation of Radiological

Effluent Technical Specifications for Units 1 and 2.

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:

Criterion 1 - Does not Involve a Significant Increase in the

probability or Consequences of an Accident Previously Evaluated.

The proposed changes are considered administrative in nature.

These changes alter only the location of programmatic controls and

procedural details relative to radioactive effluents, radiological

environmental monitoring, solid radioactive wastes, and associated

reporting requirements. Compliance with applicable regulatory

requirements will continue to be maintained. In addition, the

proposed changes do not alter the conditions and assumptions in any

of the Safety Analysis Report (SAR) accident analyses. Since the SAR

accident analyses remain bounding, the radiological consequences

previously evaluated are not adversely affected by the proposed

changes.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The proposed changes do not involve any changes to the

configuration or method of operation any plant equipment. The

proposed changes are considered administrative in nature.

Accordingly, no new failure modes have been defined for any plant

system or component important to safety nor has any new limiting

single failure have been identified as a result of the proposed

changes. Also, there will be no change in types or increase in the

amounts of any radioactive effluents released offsite.

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

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in Margin

of Safety.

The proposed changes do not involve nay actual change in the

methodology used in the control of radioactive effluents, solid

radioactive wastes, or radiological environmental monitoring. These

changes are considered administrative in nature and provide for the

relocation of procedural details outside the Technical

Specifications. This change adds appropriate administrative controls

in the Technical Specifications to provide continued assurance of

compliance with applicable regulatory requirements.

Therefore, this change does not involve a significant reduction

in the margin of safety. I21Therefore, based upon the reasoning

presented above and the previous discussion of the amendment

request, Entergy Operations has determined that the requested change

does not involve significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

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

NRC Project Director: William D. Beckner

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

Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

Date of amendment request: October 2, 1996

Description of amendment request: Relocation of Selected Technical

Specifications Instrumentation Requirements Allowed by Generic Letter

95-10

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:

Criterion 1 - Does not Involve a Significant Increase in the

probability or Consequences of an Accident Previously Evaluated.

The [Nuclear Regulatory Commission] NRC issued Generic Letter

(GL) 95-10 to allow licensees to relocate certain instrumentation

requirements to licensee controlled documents or programs. The staff

has concluded that the specifications listed in the GL were not

required to be included in the technical specifications as required

by 10 CFR 50.36. The staff concluded that the instrumentation

addressed in these specifications are not related to dominant

contributors to plant risk.

The specifications included in this amendment request are being

relocated to the Technical Requirements Manual (TRM). Once in the

TRM, future changes to these requirements will be controlled under

10 CFR 50.59. By controlling future changes under 10 CFR 50.59, NRC

review and approval will be requested for changes exceeding the

regulatory threshold of an unreviewed safety question.

This amendment request does not remove or modify any of the

instrumentation requirements for either unit. This amendment request

does not affect any of the accident initiators, conditions or

assumptions for any of the accidents previously evaluated.

Therefore, this change does not involve a significant increase in

the probability of any accident previously evaluated.

This amendment request is administrative in nature and does not

affect any system or component functional requirements. This change

does not affect the operation of the plant or affect any component

that is used to mitigate the consequences of any accident.

Therefore, this change does not involve a significant increase in

the consequences of any accident previously evaluated.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The relocation of existing requirements from the technical

specifications to other licensee controlled documents is considered

administrative in nature. This change does not modify or remove any

plant instrumentation requirements. This proposed change will not

affect any plant system or structure, nor will it affect any system

functional or operability requirements. Consequently, no new failure

modes are introduced as a result of this change. Therefore, this

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

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in Margin

of Safety.

The proposed amendment request represents a relocation of a

portion of the information previously located in each unit's

technical specification instrumentation section to other licensee

controlled documents that ate controlled under 10 CFR 50.59. The

proposed change is administrative in nature because the

instrumentation requirements for the facility remain the same.

[[Page 2189]]

The proposed change does not represent a change in the configuration

or operation of the plant. Therefore, this change does not involve a

significant reduction in the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

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

NRC Project Director: William D. Beckner

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 2, 1996

Description of amendment request: The proposed Technical

Specification (TS) Change Request will permit the use of 10 CFR Part 50

Appendix J, Option B, Performance-Based Containment Leakage Testing for

Type A, B and C leak rate testing. TSs 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3,

4.6.1.6 and 4.6.1.7 are revised and Section 6.15 is added establishing

the Containment Leakage Rate Testing Program. The Bases are revised to

reflect this change. Minor editorial changes are included in this

request. Waterford Steam Electric Station is planning to have a

Containment Leakage Rate Testing Program in place prior to the next

scheduled refueling outage. This program will be in accordance with the

guidelines contained in Regulatory Guide 1.163, ``Performance-Based

Containment Leak-Test Program,'' dated September 1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change will not affect the assumptions, design

parameters, or results of any accident previously evaluated. The

proposed change does not add or modify any existing equipment. The

proposed changes will result in increased intervals between

containment leakage tests determined through a performance based

approach. The intervals between such tests are not related to

conditions which cause accidents. The proposed changes do not

involve a change to the plant design or operation. Therefore, this

change does not involve a significant increase in the probability of

any accident previously evaluated.

NUREG-1493, ``Performance-Based Containment Leak-Test Program,''

contributed to the technical bases for Option B of 10 CFR 50

Appendix J. NUREG-1493 contains a detailed evaluation of the

expected leakage from containment and the associated consequences.

The increased risk due to lengthening of the intervals between

containment leakage tests was also evaluated and found acceptable.

Using a statistical approach, NUREG-1493 determined the increase in

the expected dose to the public from extending the testing frequency

is extremely small. It also concluded that a small increase is

justifiable due to the benefits which accrue from the interval

extension. The primary benefit is in the reduction in occupational

exposure. The reduction in the occupational exposure is a real

reduction, while the small increase to the public is statistically

derived using conservative assumptions. Therefore, this change does

not involve a significant increase in the consequences of any

accident previously evaluated.

The proposed change does not involve modifications to any

existing equipment. The proposed change will not affect the

operation of the plant or the manner in which the plant is operated.

The reduced testing frequency will not affect the testing

methodology. Therefore, the proposed change will not create the

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

previously evaluated.

The proposed change does not change the performance methodology

of the containment leakage rate testing program. However, the

proposed change does affect the frequency of containment leakage

rate testing. With an increased frequency between tests, the

proposed change does increase the probability that a increase in

leakage could go undetected for a longer period of time. Operational

experience has demonstrated the leak tightness of the containment

buildings has been significantly below the allowable leakage limit.

The margin of safety that has the potential of being impacted by

the proposed change involves the offsite dose consequences of

postulated accidents which are directly related to containment

leakage rates. The limitation on containment leakage rate is

designed to ensure the total leakage volume will not exceed the

value assumed in our accident analysis. The margin of safety for the

offsite dose consequences of postulated accidents directly related

to containment leakage is maintained by meeting the 1.0 La

acceptance criteria. The proposed change maintains the 1.0 La

acceptance criteria. 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: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant

Unit 1, St. Lucie County, Florida

Dates of amendment request: December 9, 1996

Description of amendment request: The licensee proposed to modify

specifications for selected cycle-specific reactor physics parameters

to refer to the St. Lucie Unit 1 Core Operating Limits Report (COLR)

for limiting values. Minor administrative changes are also included.

The proposed Technical Specification (TS) changes utilized the guidance

provided in Generic Letter 88-16 and are intended to be consistent with

the Standard Technical Specifications for Combustion Engineering Plants

(NUREG-1432, 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) 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.

The proposed amendment relocates the calculated values of

selected cycle-specific reactor physics parameter limits from the TS

to the COLR, and includes minor editorial changes which do not alter

the intent of stated requirements. The amendment is administrative

in nature and has no impact on any plant configuration or system

performance relied upon to mitigate the

consequences of an accident. Parameter limits specified in the

COLR for this amendment are not changed from the values presently

required by Technical Specifications. Future changes to the

calculated values of such limits may only be made using NRC approved

methodologies, must be consistent with all applicable safety

analysis limits, and are controlled by the 10 CFR 50.59 process.

Assumptions used for accident initiators and/or safety analysis

acceptance criteria are not changed by this amendment. Therefore,

operation of the facility in accordance with the proposed amendment

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

[[Page 2190]]

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment relocates the calculated values of cycle

specific reactor physics limiting parameters to the COLR and will

not change the physical plant or the modes of operation defined in

the facility license. The changes do not involve the addition of new

equipment or the modification of existing equipment, nor do they

alter the design configuration of St. Lucie plant systems.

Therefore, operation of the facility in accordance with the 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 cycle specific parameter limits being relocated to the COLR

by this amendment have not been changed from the values presently

required by the TS, and a requirement to operate the plant within

the bounds of the limits specified in the COLR is retained in the

individual specifications. Future changes to the calculated values

of these limits by the licensee may only be developed using NRC-

approved methodologies, must remain consistent with all plant safety

analysis limits addressed in the Final Safety Analysis Report

(FSAR), and are further controlled by the 10 CFR 50.59 process. As

discussed in Generic Letter 88-16, the administrative controls

established for the values of cycle specific parameters using the

guidance of that letter assure conformance with 10 CFR 50.36. Safety

analysis acceptance criteria are not being altered by this

amendment. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a significant reduction in

a margin of safety.

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

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, Florida 33408

NRC Project Director: Frederick J. Hebdon

Florida Power and Light Company, Docket No. 50-335 St. Lucie Plant

Unit 1, St Lucie County, Florida

Date of amendment request: December 20, 1996

Description of amendment request: The licensee proposed to delete a

footnote associated with TS 2.1.1, ``Reactor Core Safety Limits,''

which requires reactor thermal power to be limited to 90% of 2700

Megawatts thermal for Cycle 14 operation beyond 7000 Effective Full

Power Hours [EFPH]. The thermal power limit was required pending

completion of a Small Break Loss of Coolant Accident (SBLOCA)

reanalysis that demonstrated acceptable results using input assumptions

corresponding to an increased number of steam generator tubes being

plugged. The SBLOCA reanalysis was completed and included with the

submittal.

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.

The proposed change will allow full Cycle 14 operation at 100%

of rated power (2700 MWth), by deleting the requirement to

derate to 90% of rated power prior to exceeding 7000 EFPH. This

restriction was imposed in the NRC transmittal letter for License

Amendment 145 for SBLOCA considerations when considering the

increased SGTP [steam generator tube plugging]

level of 30% plus or minus 7%. All Final Safety Analysis Report

(FSAR) events, other than SBLOCA were evaluated at 100% of rated

thermal power and showed no significant increases in the probability

or consequences of accidents previously evaluated.

The SBLOCA was reanalyzed to demonstrate continued compliance

with 10 CFR 50.46 criteria. There is no impact of the proposed

change on any FSAR accident initiator. The plant configuration and

systems remain unchanged.

Therefore, 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.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

This proposed amendment removes the requirement in the Technical

Specifications to derate to 90% of 2700 MWth for Cycle 14 operation

beyond 7000 EFPH. There will be no change to the modes of operation

of the plant. The plant configuration and the design functions of

all the safety systems remain unchanged.

The proposed amendment will not change the physical plant or the

modes of operation defined in the facility license. The changes do

not involve the addition of new equipment or the modification of

existing equipment, nor do they alter the design of St. Lucie plant

systems. Therefore, operation of the facility in accordance with the

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 impact of the proposed change on available margin to the

acceptance criteria for Specified Acceptable Fuel Design Limits

(SAFDL), primary and secondary over-pressurization, peak containment

pressure, potential radioactive releases, 10 CFR 50.46 requirements

for the large break LOCA, and existing limiting conditions for

operation has been evaluated and addressed in the reduced RCS

[reactor coolant system] flow operating license Amendment No. 145. A

requirement to derate to 90% of 2700 MWth was imposed based on the

SBLOCA analysis. The small break LOCA analysis with 30% plus or

minus 7% SGTP

supported operation up to 7000 EFPH at 100% of rated thermal

power. A reanalysis of SBLOCA with the limiting end-of-cycle

conditions at 100% of rated power, demonstrates continued compliance

with 10 CFR 50.46 criteria.

Therefore, operation of the facility in accordance with the

proposed amendment would not involve a significant reduction in a

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

and, based on 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, Florida 33408

NRC Project Director: Frederick J. Hebdon

GPU Nuclear Corporation, Docket No. 50-289, Three Mile Island, Unit

1, Dauphine County, Pennsylvania

Date of amendment request: December 3, 1996

Description of amendment request: This amendment will incorporate

certain improvements from the Standard Technical Specifications for B&W

Plants (NUREG-1430).

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:

GPU Nuclear has determined that this Technical Specification

Change Request involves no significant hazards consideration as

defined in 10 CFR 50.92 because:

1. Operation of the facility in accordance with the proposed

amendment would not

[[Page 2191]]

involve a significant increase in the probability of occurrence or

the consequences of an accident previously evaluated. The proposed

amendment deletes limiting condition for operation (LCOs) from the

TMI-1 Technical Specifications that are no longer required to be

addressed in Technical Specifications per 10 CFR 50.36(c)(2)(ii).

The proposed amendment also deletes a Surveillance requirement from

the TMI-1 Technical Specifications. This surveillance requirement

has no corresponding LCO and is formatted in the typical LCO format.

These items are addressed in licensee controlled documents. This

proposed amendment incorporates relaxation of selected timeclocks

and surveillances frequencies consistent with NUREG 1430 and adds a

timeclock to a unique LCO. The proposed changes do not modify the

operation, limits or controls of systems, structures or components

relied upon to prevent or mitigate the consequences [of] accidents

previously evaluated. Also, the reliability of systems and

components relied upon to prevent or mitigate the consequences of

accidents previously evaluated is not degraded by the proposed

changes. Therefore, this change does not involve a significant

increase in the probability of occurrence or the 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

kind of accident from any accident previously evaluated because no

new failure modes are created by the proposed changes.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment does not change any operating limits

for reactor operation.

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 staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

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

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

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

NRC Project Director: John F. Stolz

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: October 25, 1996

Description of amendment requests: The proposed amendments would

incorporate the requirements of 10 CFR Part 50, Appendix J, Option B

for containment leakage tests. In addition, the amendments would add a

new section to Technical Specifications, which establishes the

requirements of the containment leakage rate testing program,

consistent with the Improved Standard Technical 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. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes provide a mechanism within the TS for

implementing a performance-based leakage rate test program which was

promulgated by the revision to 10 CFR Part 50 to incorporate Option

B to Appendix J. The proposed changes do not involve any physical or

operational changes to structures, systems or components. The

current safety analyses and safety design basis for the accident

mitigation functions of the containment, the airlocks, and the

containment isolation valves are maintained. Since the allowable

containment leakage is still maintained within the analyzed limit

assumed in the accident analysis, there is no adverse effect on

either onsite or offsite dose consequences. Therefore, these changes

will not increase the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

analyzed.

The proposed changes do not involve any physical or operational

changes to structures, systems or components. No new failure

mechanisms beyond those already considered in the current plant

safety analyses are introduced. Therefore, the proposed changes do

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

from any accident previously analyzed.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

Extending containment leakage rate test intervals from those

currently provided in the Technical Specifications to those provided

for in 10 CFR (Part) 50 Appendix J, Option B may slightly increase

the risk due to an increased likelihood of containment leakage

corresponding to the increased testing intervals. However, this is

somewhat compensated by the corresponding risk reduction benefits

received from the reduction in component cycling, stress, and wear

associated with the increased intervals. When considering the total

integrated risk, which includes all analyzed accident sequences, the

possible additional risk associated with increasing test intervals

is negligible.

The NRC letter to NEI (Nuclear Energy Institute) dated November

2, 1995, recognizes that changes similar to the proposed changes at

PINGP (Prairie Island Nuclear Generating Plant) are required to

implement Option B of 10 CFR (Part) 50, Appendix J. In NUREG-1493,

``Performance-Based Containment Leak-Test Program'', dated September

1995, which forms the basis for the Appendix J revision, the NRC

concludes that adoption of performance-based testing will not

significantly reduce the margin of safety. The containment leak rate

data and component performance history at PINGP are consistent with

the conclusions reached in NUREG-1493 and NEI 94-01. Thus, the

proposed license amendments do not involve a significant reduction

in a margin of safety and will continue to support the regulatory

goal of ensuring an essentially leak-tight containment boundary.

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: November 18, 1996

Description of amendment request: The amendments would amend the

Technical Specifications for Susquehanna Steam Electric Station (SSES),

Units 1 and 2 by increasing the maximum isolation times for the reactor

core isolation cooling inboard warm-up line isolation valves (HV129F088

and HV249F088) from 3 seconds to 12 seconds, the high pressure core

[[Page 2192]]

injection inboard warm-up line isolation valves (HV-155F100 and HV-

255F100) from 3 seconds to 6 seconds and the reactor recirculation

process sample line (RRPSL) isolation valves (HV143F019 and HV243F019)

from 2 seconds to 9 seconds.

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.

Chapters 6,9, and 15 of the FSAR [final safety analysis report],

current operating cycles Reload Summary Reports for Units 1 and 2,

Design Basis Document DBD046 (Seismic and Hydrodynamic Loads), and

NUREG-0776 (Safety Evaluation Report for SSES), were reviewed to

determine if the proposed action has an effect on the spectrum of

analyzed anticipated operational transients or postulated design

basis accidents.

The proposed modifications involve replacing the pilot solenoid

valves on the Reactor Recirculation Loop ``B'' Process Sample Line

Isolation Valve (HV1/243F019) and the inboard RCIC [reactor core

isolation cooling] and HPCI [high pressure core injection] Steam

Warm-Up Line Isolation Valves (HV-1/249F088 and HV-1/255F100). They

do not alter any system operation or control logic other than to

increase the time it takes for the associated containment isolation

valve to close. As discussed above, the effects of the increased

isolation times for RCIC and HPCI impacted lines are bounded by the

larger parallel lines with isolation times much greater than the new

isolation times for the smaller lines. In the case of the Reactor

Recirculation Loop ``B'' Process Sample Line, the worst case

scenario for a line of that size is addressed in FSAR Section 15.6.2

and the results have been found acceptable. In fact, the line

breakage event analyzed in the FSAR section postulates a break

outside containment that is not isolable and that does not require

operator action for up to 10 minutes.

The modifications enhance isolation valve performance by

ensuring proper operation in the event of a degraded air system.

Failures within the Process Sampling, RCIC or HPCI systems or

their components are not postulated as causes of accident scenarios

nor is increasing the stroke time of the subject containment

isolation valves [HV-1/243F019]. These systems provide safety

features utilized to mitigate the consequences of the accidents.

However, the failure mode of the replacement solenoid valve is

similar in each case to that of the solenoid valve being replaced in

that it closes upon loss of power or loss of air supply. The current

ability of the plant design to meet the single failure criterion is

unchanged by this modification.

Based on the above discussion, the proposed action does not

involve a significant increase in the probability or consequences of

an accident as previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

Chapters 6, 9, and 15 of the FSAR were reviewed to determine if

the proposed action [valve replacement with increased isolation

times for associated HPCI, RCIC, RRPSL valves] has the potential of

creating a postulated initiating event which is different than the

analyzed anticipated operational transients or postulated design

basis accident addressed. The review did not identify a postulated

initiating event which would create the possibility for an accident

of a different type due to replacing the pilot solenoid valves of

the affected Reactor Recirculation LOOP ``B'' Process Sample Line or

RCIC or HPCI Steam Warm-Up Line isolation valves.

Also, the Reactor Recirculation Process Sample Line, as part of

the Process Sampling System described in FSAR section 9.3.2.3, does

not perform any safety functions. It is simply an alternate means

for in line reactor water chemistry monitoring upon the loss of the

RWCU system, and its loss does not create any possibility for

unevaluated accidents or malfunctions.

Thus, replacing the pilot solenoid valves on the affected

Reactor Recirculation Process Sample Line, RCIC Steam Warm-Up Line,

and HPCI Steam Warm-Up Line isolation valves as well as relocating

the Process Sample Line solenoid valve for EQ [equipment

qualification] purposes does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed action involves replacing existing pilot solenoid

valves on containment isolation valves for the Process Sampling,

RCIC, and HPCI Systems, as listed above, with direct acting solenoid

valves to ensure proper valve operation in the event of a degraded

air or gas system as well as relocating the Process Sampling pilot

solenoid valve for EQ purposes.

a. Reactor Recirculation Loop ``B'' Process Sample Line

The limiting condition for the operation of the Reactor

Recirculation Loop ``B'' Process Sample Line Inboard Isolation Valve

(HV-1/243F019) is governed by Technical Specification Section 3/

4.6.3 and its Bases which presently requires this valve to close

within 2 seconds as defined in Technical Specification Table 3.6.3-

1. The proposed modifications involve replacing the pilot solenoid

valve of the normally open isolation valve (HV-1/243F019) with a

direct acting pilot solenoid valve as well as relocating the pilot

solenoid valve to assure an EQ life which supports a 24 month

operating cycle. The combined effects of a lower flow coefficient

and relocating the solenoid valve will require an increase in the

Technical Specification Table 3.6.3-1 isolation time from 2 seconds

to 9 seconds.

This increase in isolation time does not reduce the margin of

safety as defined in the Technical Specification Section Basis,

because breakage of lines of this size is addressed in the

Susquehanna SES [steam electric station] FSAR Section 15.6.2 and the

results found acceptable. In fact, the line breakage event analyzed

postulates a break outside containment that is not isolable and that

does not require operator action for up to 10 minutes. Also, it is

noted that the outboard isolation valve, HV-1/243F020, also closes

on the same containment isolation signal, and its Technical

Specification isolation time limit remains 2 seconds.

The failure mode of the affected Reactor Recirculation Loop

``B'' Process Sample Line Inboard isolation valve is to close on

loss of power or air supply, therefore, the proposed modifications

do not affect the operability of the isolation valve or reduce the

margin of safety.

b. RCIC

The limiting condition for operation of the RCIC system is

governed by Technical Specification Section 3/4.7.3 and its Bases

which requires RCIC to be operable as the primary non-ECCS source of

emergency core cooling. The proposed modifications involve replacing

the pilot solenoid valve of the normally closed Steam Warm-Up Line

Isolation Valve (HV-1/249F088). This valve can be manually opened in

the absence of an isolation signal to permit steam from the reactor

to pressurize and warm the steam supply line downstream of the HV-1/

249F007 valve.

Installation of the direct acting solenoid valve will require an

increase in the Technical Specification Section 3/4.6.3 isolation

time for the RCIC Steam Warm-Up Line Isolation Valve (HV-1/249F088)

from 3 seconds to 12 seconds but does not reduce the margin of

safety as defined in the Technical Specification Section Basis. The

increase in closure time for the HV-1/249F088 isolation valve does

not compromise the overall line isolation due to the fact that the

impact of these 1'' warm up line valves is enveloped by the impact

of the much larger 4'' RCIC inboard and outboard isolation valves

(HV-1/249F007 and HV-1/249F008), which remain open an additional 8

seconds before isolating. The 4'' valves are the limiting components

for providing containment isolation for this line.

The failure mode of the affected RCIC Steam Warm-Up Line

Isolation Valve is to close, if open, on loss of power or air

supply, therefore, the proposed modifications do not affect the

operability of the isolation valve or reduce the margin of safety.

c. HPCI

The limiting condition for operation of the HPCI system is

governed by Technical Specification Section 3/4.5.1 and its Bases

which requires HPCI to be operable for proper Emergency Core Cooling

System operation. Operability includes the HPCI pump and a flow path

capable of taking suction from the suppression pool and delivering

the water to the reactor vessel. The proposed modifications involve

replacing the pilot solenoid valve of the normally closed Steam

Warm-Up Line Isolation Valve (HV-1/255F100). This valve can be

manually opened in the absence of an isolation signal, to permit

steam from the reactor to pressurize

[[Page 2193]]

and warm the steam supply line downstream of the HV-1/255F002 valve.

Installation of the direct acting solenoid valve will require an

increase in the Technical Specification Section 3/4.6.3 isolation

time for the HPCI Steam Warm-Up Line Isolation Valve (HV-1/255F100)

from 3 seconds to 6 seconds but does not reduce the margin of safety

as defined in the Technical Specification Section Basis. The

increase in closure time for the HV-1/255F100 isolation valve does

not compromise the overall line isolation due to the fact that the

impact of these 1'' warm up line valves is enveloped by the impact

of the much larger 10'' HPCI inboard and outboard isolation valves

(HV-1/255F002 and HV-1/255F003) which remain open an additional 44

seconds before isolating. The 10'' valves are the limiting

components for providing containment isolation for this line.

The failure mode of the affected HPCI Steam Warm-Up Line

Isolation Valve is to close, if open, on loss of power or air

supply, therefore, the proposed modifications do not affect the

operability of the isolation valve or reduce the margin of safety.

Thus, based on a review of the Technical Specification, their

Bases, the FSAR and NUREG 0776 (Safety Evaluation Report for SSES),

the replacement of the pilot solenoid valves 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: 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

Pennsylvania Power and Light Company, Docket No. 50-388,

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania

Date of amendment request: December 18, 1996

Description of amendment request: The amendment would change the

Susquehanna Steam Electric Station Unit 2 Technical Specifications to

reflect the use of a 24-month operating cycle and the use of the

ATRIUM-10 fuel design. The amendment includes changes to two

definitions in Section 1, inclusion of new minimum critical power ratio

safety limits in Sections 2.1.2 and 3.4.1.1.2, changes in Section 5.3.1

to reflect the new fuel design, and the listing of Siemens Power

Corporation topical reports in Section 6.9.3.2.

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 applicable sections of the FSAR [Final Safety Analysis

Report] are Chapters 5,6.3,9, and 15 of the FSAR. Chapter 5

discusses the results of the ASME overpressure analyses for the

reactor pressure boundary. Chapter 6.3 discusses the LOCA [loss-of-

coolant accident]. Chapter 9 discusses fuel storage and handling.

Chapter 15 describes the transient and accident analyses, a majority

of which have been generically dispositioned to be non-limiting. A

discussion of the impact of the Technical Specification changes is

provided below.

The change to Definitions 1.2 and 1.3 makes the definitions

applicable to ATRIUM-10. There are no effects on safety functions

from this change.

A cycle specific MCPR [minimum critical power ratio] Safety

Limit analysis was performed for PP&L [Pennsylvania Power & Light

Company] by SPC [Siemens Power Corporation]. This analysis used NRC

approved methods described in Technical Specification Reference 13

(ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2.). The SAFETY

LIMIT MCPR calculation statistically combines uncertainties on

feedwater flow, feedwater temperature, core flow, core pressure,

core power distribution, and the uncertainty in the Critical Power

Correlation. The SPC analysis used cycle specific power

distributions and calculated MCPR values such that at least 99.9% of

the fuel rods are expected to avoid boiling transition during normal

operation or anticipated operational occurrences. The resulting two-

loop and single-loop values (Technical Specification sections 2.1.2

and 3.4.1.1.2) are included in the proposed change. Thus, the

cladding integrity and its ability to contain fission products is

not adversely affected.

The change to the Design Features (Section 5.3) increases the

allowable enrichment. Analyses have demonstrated that the ATRIUM-10

fuel will remain subcritical (k-effectiveavg

greater than or equal to 280 deg.F. Initiating conditions and

assumptions remain as previously analyzed for accidents in the DBNPS

Updated Safety Analysis Report.

These revisions do not involve any physical changes to systems

or components, nor do they alter the typical manner in which the

systems or components are operated.

Review results of historical 18 month surveillance data and

maintenance records support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30 months on a non-routine

basis) because little, if any, potential for an increase in a

failure rate of a system or component was identified during these

reviews.

These proposed revisions are consistent with NRC guidance on

evaluating and proposing such revisions as provided in Generic

Letter 91-04, ``Changes in Technical Specification Surveillance

Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2,

1991.

The proposed revision to Technical Specification Table 1.2,

Frequency Notation, and the related proposed revision from an

[[Page 2195]]

``R'' frequency notation to an ``E'' frequency notation for

Technical Specification Surveillance Requirements that are remaining

on an 18 month frequency, are administrative in nature, do not

change current actual Technical Specification requirements, and do

not affect previously evaluated accidents.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because the source term, containment

isolation or radiological releases are not being changed by these

proposed revisions. Existing system and component redundancy is not

being changed by these proposed changes. Existing system and

component operation is not being changed by these proposed changes.

The assumptions used in evaluating the radiological consequences in

the DBNPS Updated Safety Analysis Report are not invalidated.

The proposed revision to Technical Specification Table 1.2,

Frequency Notation, and the related proposed revision from an ``R''

frequency notation to an ``E'' frequency notation for Technical

Specification Surveillance Requirements that are remaining on an 18

month frequency, are administrative in nature, do not change current

actual Technical Specification requirements, and do not affect

previously evaluated accidents.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because these

revisions do not involve any physical changes to systems or

components, nor do they alter the typical manner in which the

systems or components are operated.

Review results of historical 18 month surveillance data and

maintenance records support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30 months on a non-routine

basis) because little, if any, potential for an increase in a

failure rate of a system or component was identified during these

reviews. No changes are being proposed to the type of testing being

performed, only to the length of the surveillance test interval.

The proposed revision to Technical Specification Table 1.2,

Frequency Notation, and the related proposed revision from an ``R''

frequency notation to an ``E'' frequency notation for Technical

Specification Surveillance Requirements that are remaining on an 18

month frequency, are administrative in nature, do not change current

actual Technical Specification requirements, and do not affect the

manner in which systems and components are being operated or tested.

3. Not involve a significant reduction in a margin of safety

because the review results of the historical 18 month surveillance

data and maintenance records identified little, if any, potential

for an increase in a failure rate of a system or component due to

increasing the surveillance test interval to 24 months. Existing

system and component redundancy is not being changed by these

proposed changes.

The proposed revision to Technical Specification Table 1.2,

Frequency Notation, and the related proposed revision from an ``R''

frequency notation to an ``E'' frequency notation for Technical

Specification Surveillance Requirements that are remaining on an 18

month frequency, are administrative in nature, do not change current

actual Technical Specification requirements, and do not reduce the

margin of safety.

There are no new or significant changes to the initial

conditions contributing to accident severity or consequences,

therefore there are no significant reductions 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: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

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

supplementedNovember 26, 1996.

Brief description of amendment: The proposed amendment will modify

Table 3.1.1, ``Reactor Protection System (SCRAM) Instrumentation

Requirement,'' Table 3.2.C.1, ``Instrumentation That Initiates Rod

Blacks,'' and Technical Specification 3/4.4, ``Standby Liquid

Control.''

Date of issuance: December 27, 1996

Effective date: December 27, 1996

Amendment No.: 169

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

Technical Specifications.

Date of initial notice in Federal Register: June 6, 1995 (61 FR

28606) The November 26, 1996, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration.The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 27, 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: July 26, 1996, as supplemented

on September 3, 1996, September 18, 1996, two submittals dated October

14, 1996, October 22, 1996, two submittals dated November 8, 1996, and

December 17, 1996.

Brief description of amendments: The amendments allow Commonwealth

Edison Company to control the reactor coolant system pressure and

temperature limits for heatup, cooldown, low temperature operation and

hydrostatic testing. They also revise the reactor vessel material

surveillance program specimen withdrawal schedule

[[Page 2196]]

such that the Unit 2 removal of capsule X is delayed until 19 Effective

Full Power Years.

Date of issuance: December 20, 1996

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 177 and 164

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

revised the Technical Specifications.

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

FR 50341). The September 3, 1996, September 18, 1996, two submittals

dated October 14, 1996, October 22, 1996, two November 8, 1996, and

December 17, 1996, submittals provided additional clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

December 20, 1996.No significant hazards consideration comments

received: No

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe

County, Michigan

Date of application for amendment: March 25, 1996 (NRC-96-0003)

Brief description of amendment: The amendment revises the testing

requirements used to determine the operability of the charcoal in the

engineered safety feature systems.

Date of issuance: December 23, 1996

Effective date: December 23, 1996, with full implementation within

45 days

Amendment No.: 110

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 31, 1996 (61 FR

40014) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 23, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

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

Date of application for amendments: December 11, 1996, as

supplemented December 17, 19, and 26, 1996

Brief description of amendments: The amendments approve changes to

the Updated Final Analysis Report (UFSAR), and require that the changes

be submitted with the next update of the UFSAR pursuant to 10 CFR

50.71(e). The associated Safety Evaluation delineates the staff's

review and findings regarding the one-time emergency power engineered

safeguards functional test.

Date of issuance: January 2, 1997

Effective date: January 2, 1997

Amendment Nos.: 220, 220, 217

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

amendments revised the Updated Final Safety Analysis Report. Public

comments requested as to proposed no significant hazards consideration:

Yes. (61 FR 66699 December 18, 1996) The notice provided an opportunity

to submit comments on the Commission's proposed no significant hazards

consideration determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by January 2,

1997, as corrected to read January 17, 1997, but indicated that if the

Commission makes a final no significant hazards consideration

determination, any such hearing would take place after issuance of the

amendments.

The December 17, 19, and 26, 1996, letters provided additional

information that did not change the scope of the December 11, 1996,

application and initial proposed no significant hazards consideration

determination.

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

exigent circumstances, and a final no significant hazards consideration

determination are contained in a Safety Evaluation dated January 2,

1997.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 31, 1996

Brief description of amendment: The amendment revises the technical

specifications to increase the amount of trisodium phosphate (TSP)

dodecahydrate located in the containment sump storage baskets.

Date of issuance: December 30, 1996

Effective date: December 30, 1996

Amendment No.: 179

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 31, 1996 (61 FR

40025) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 30, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Philadelphia Electric Company, Docket No. 50-353, Limerick

Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: August 1, 1996

Brief description of amendment: This amendment revised the

Technical Specifications Section 3/4.4.6 (i.e., Figure 3.4.6.1-1) to

reflect the addition of two hydrotest curves, effective for 6.5 and 8.5

Effective Full Power Years (EFPY), to the existing Pressure-Temperature

Operating Limit (PTOL) curves for LGS Unit 2.

Date of issuance: December 30, 1996

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment No.: 80

Facility Operating License No. NPF-85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 6, 1996 (61 FR

57490) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 30, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Notice Of Issuance Of Amendment To Facility Operating License And

FinalNo Significant Hazards Consideration Determination

During the period since publication of the last biweekly notice,

individual notices of issuance of amendments have been issued for the

facilities as listed below. These notices were previously published as

separate individual notices. They are repeated here because this

biweekly notice lists all amendments that have been issued for which

the Commission has made a final determination that an amendment

involves no significant hazards consideration.

In this case, a prior Notice of Consideration of Issuance of

Amendment, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing was issued, a hearing was requested, and

the amendment was issued before any hearing because the Commission made

a final determination that the

[[Page 2197]]

amendment involves no significant hazards consideration.

Details are contained in the individual notice as cited.

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

Limerick Generating Station, Unit 2, Montgomery County,

Pennsylvania

Date of amendment request: December 6, 1996

Brief description of amendment request: The amendment would revise

Technical Specification (TS) Section 2.1 and its associated TS Basis to

reflect the change in the Minimum Critical Power Ratio Safety Limit due

to the use of GE13 fuel product line and the cycle-specific analysis

performed by General Electric Company (GE), for Limerick Generating

Station, Unit 2, Cycle 5.

Date of publication of individual notice in Federal Register:

December 23, 1996 (61 FR 67582)

Expiration date of individual notice: January 22, 1997

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Dated at Rockville, Maryland, this 8th day of January 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

Director, Division of Reactor Projects - III/IV, Office of Nuclear

Reactor Regulation

[Doc. 97-848 Filed 1-14-97; 8:45 am]

BILLING CODE 7590-01-F