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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 22, 1996, through December 6, 1996.

The last biweekly notice was published on December 4, 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 17, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

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

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one

[[Page 66703]]

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. 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: September 20, 1996.

Description of amendment request: The proposed amendments would

update the Pressure Temperature (P-T) curves contained in the Technical

Specifications to 22 Effective Full Power Years (EFPYs).

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:

The proposed changes merely adjust the reference temperature for

the limiting beltline material to account for irradiation effects

and provide the same level of protection as previously evaluated.

The adjusted reference temperature calculations were performed

utilizing the guidance contained in Regulatory Guide 1.99, Revision

2. The change is administrative in nature to reflect the extension

of the operating limits to 22 EFPY. As such, these changes will not

significantly increase the probability or consequences of a

previously evaluated accident.

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

from any accident previously evaluated because:

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

different kind of accident previously evaluated for Dresden or Quad

Cities Stations. No new modes of operation are introduced by the

proposed changes. The revised operating limits are merely an update

of the old limits by taking into account the effects of irradiation

on the limiting reactor vessel material. Use of the revised P-T

curves will continue to provide the same level of protection as was

previously reviewed and approved. Therefore, the proposed changes do

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

from any previously evaluated.

The associated change to the P-T curves related to this proposed

amendment does not affect any activities or equipment and are not

assumed in any safety analysis to initiate any accident sequence for

Dresden or Quad Cities Stations; therefore, the proposed changes do

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

from any previously evaluated.

3. Involve a significant reduction in the margin of safety

because:

The proposed amendment reflects an update of the P-T curves to

extend the operating limit to 22 EFPY. The revised curves are based

on the latest NRC guidance along with actual data for the units. The

new limits retain the margin of safety to the level expected for a

new vessel, adjusted for irradiation effects as required by 10 CFR,

Appendix G, thereby maintaining a conservative margin of safety.

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.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 31, 1996.

Description of amendment request: The proposed amendments would

relocate the requirements for seismic monitoring instrumentation from

the Technical Specifications to licensee controlled documents. The

Technical Specifications affected are 3/4.3.7.2, ``Seismic Monitoring

Instrumentation,'' Table 3.3.7.2-1, ``Seismic Monitoring

Instrumentation,'' Table 4.3.7.2-1, ``Seismic Monitoring

Instrumentation Surveillance Requirements,'' and Bases Section 3/

4.3.7.2, ``Seismic Monitoring 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:

[[Page 66704]]

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The function of the seismic monitoring instrumentation is to

monitor seismic activity above the Operating-Basis Earthquake (OBE)

threshold, and to record observed seismic data for comparison to

design basis response spectra. The seismic monitoring

instrumentation does not provide any function to mitigate an

accident or the consequences of an accident. The replacement seismic

monitoring instrumentation will remain in place. The proposed

Amendment is not a result of any changes to system function, alarm

setpoints, or main control room annunciators. Rather, the Technical

Specification requirements (as revised for the replacement

instrumentation) are being relocated to licensee-controlled

documents in accordance with NRC Generic Letter 95-10.

The proposed change relocates requirements and surveillances for

structures, systems, components or variables that do not meet the

criteria for inclusion in Technical Specifications as identified in

the Application of Selection Criteria to the LaSalle Technical

Specifications. The affected structures, systems, components or

variables are not assumed to be initiators of analyzed events and

are not assumed to mitigate accident or transient events. The

requirements and surveillances for these affected structures,

systems, components or variables will be relocated from the

Technical Specifications to an appropriate administratively

controlled document which will be maintained pursuant to 10 CFR

50.59. In addition, the affected structures, systems, components or

variables are addressed in existing surveillance procedures which

are also controlled by 10 CFR 50.59 and subject to the change

control provisions imposed by plant administrative procedures, which

endorse applicable regulations and standards. Therefore, this change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

accident from any accident previously evaluated because:

The seismic monitoring instrumentation does not provide any

function to mitigate an accident or the consequences of an accident.

The replacement seismic monitoring instrumentation will remain in

place and will provide the same basic function as the existing

instrumentation. The replacement instrumentation will provide

enhanced system reliability and will not result in any changes to

system function, alarm setpoints, or main control room annunciators.

The Technical Specification requirements (as revised for the

replacement instrumentation) are being relocated to licensee-

controlled documents in accordance with NRC Generic Letter 95-10.

The proposed change does not involve any change in the methods

governing normal plant operation. The proposed change will not

impose or eliminate any requirements and adequate control of

existing requirements will be maintained. Thus, this change does not

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

any accident previously evaluated.

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

because:

The replacement seismic monitoring instrumentation will have no

impact on margin of safety. The intended function of the seismic

monitoring instrumentation, i.e. to record observed seismic data for

analysis to determine the impact on plant components, will be made

more reliable by this modification. The Technical Specification

requirements (as revised for the replacement instrumentation) are

being relocated to licensee-controlled documents in accordance with

NRC Generic Letter 95-10.

The proposed change will not reduce a margin of safety because

it has no impact on any safety analysis assumptions. In addition,

the relocated requirements and surveillances for the affected

structure, system, component or variable continue to meet the same

requirements as the existing Technical Specifications. However, the

LCO requirement specified in Section 3.3.7.2.a (to prepare and

submit a Special Report to the NRC within 10 days of the seismic

monitoring instrumentation being inoperable for more than 30 days)

will not be included in the ATR [Administrative Technical

Requirements] since the Technical Specification Special Report

requirements are only applicable to the LCOs. Since any future

changes to these requirements or the surveillance procedures will be

evaluated per the requirements of 10 CFR 50.59, no reduction in a

margin of safety will be permitted.

The existing requirement for NRC review and approval of

revisions, in accordance with 10 CFR 50.92, to these details

proposed for relocation does not have a specific margin of safety

upon which to evaluate. However, since the proposed change is

consistent with the BWR Standard Technical Specification, NUREG-

1434, Rev. 1 approved by the NRC Staff, revising the Technical

Specifications to reflect the approved level of detail ensures no

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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. 50-295 and 50-304, Zion

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

Date of amendment request: November 7, 1996.

Description of amendment request: The proposed amendments would

change Specification 4.3.1.A.4.b from verifying greater than or equal

to 17 percent steam generator secondary side wide range water level to

greater than or equal to 17 percent steam generator secondary side

narrow range water level.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of occurrence of any accident

previously evaluated.

Maintaining secondary side steam generator water level greater

than or equal to 17 percent by wide range level indication is the

current requirement by the technical specifications. By revising the

requirement to require using the narrow range water level, no change

in operating practices or plant configuration is made. The minimum

requirement of 17 percent by narrow range level indication is more

restrictive and conservative than 17 percent by wide range

indication. The requirement to maintain secondary side steam

generator water level greater than or equal to 17 percent by narrow

range indication is currently required by operations procedure PT-O,

Appendix F-1 and will be maintained. This change ensures that the

requirements for natural circulation cooldown are maintained in Mode

4. Therefore, changing this surveillance requirement does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not require a physical alteration of the

plant (no new or different equipment will be installed). The

Technical Specifications will continue to require OPERABLE steam

generator(s) for heat removal functions. The Technical

Specifications will continue to require the performance of SR

4.3.1.A.4.b. Changing the SR to use narrow level indication

correctly states the steam generator water level required to support

heat removal function. Thus, this change does not create the

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not result in a significant reduction in

a margin of safety because it has no impact on any safety analysis

assumptions. The requirement to have OPERABLE steam generator(s) in

MODE 4 for heat removal function is maintained. The requirement to

perform SR 4.3.1.A.4.b is not changed. Changing the SR to use narrow

level indication correctly states the steam generator water level

required to support heat

[[Page 66705]]

removal function. Therefore, this change does not involve a

significant reduction in a margin of safety.

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

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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. 50-295 and 50-304, Zion

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

Date of amendment request: November 7, 1996.

Description of amendment request: The proposed amendments would

change the values for the reduced power range neutron flux high

setpoint trip that are specified when one or more code main steam

safety valves are inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of occurrence of any accident

previously evaluated.

The requirement to change the Power Range Neutron Flux High Trip

setpoints to the Reduced Setpoint Values of Table 3.7-1 for the most

restrictive loop if one or more code MSSVs are inoperable is not

changed by this amendment. As such, no change in operating practices

or plant configuration is being made.

The amendment provides new reduced setpoint values for the Power

Range Neutron flux High Trip to ensure that for the limiting

transient (Loss of Load/Turbine Trip [LOL/TT]), a secondary side

overpressurization condition does not occur. The new values were the

result of calculation using an algorithm provided by Westinghouse in

Westinghouse Nuclear Safety Advisory Letter NSAL-94-001, ``Operation

at Reduced Power Levels with Inoperable MSSVs,'' January 25, 1994.

The new values are much more restrictive than the previous values

and ensure that the probability or consequences of an accident

previously evaluated is not increased. Therefore, the new reduced

setpoint values do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not require a physical alteration of

the plant (no new or different equipment will be installed to

implement this change). The Reduced Neutron Flux High Trip setpoints

ensure that a secondary side overpressurization transient does not

occur for the most limiting transient. In addition, no new modes of

operations will be introduced by this change. Thus, this change does

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

from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

This amendment provides new Reduced Power Range Neutron Flux

High Trip setpoints.The Specification that requires the Power Range

Neutron Flux High Trip setpoints be changed to the reduced values

for one or more inoperable MSSVs is not changed. The reduced Trip

setpoints are the result of new calculations using an algorithm

provided by Westinghouse in Westinghouse Nuclear Safety Advisory

Letter NSAL-94-001, ``Operation at Reduced Power levels with

Inoperable MSSVs,'' January 25, 1994, and ensure the LOL/TT

transient does not result in a secondary overpressurization.

Therefore, this change does not involve a significant reduction in a

margin of safety.

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

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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. 50-295 and 50-304, Zion

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

Date of amendment request: November 7, 1996.

Description of amendment request: The proposed amendments would

clarify the operability requirements for the residual heat removal

(RHR) loops during core alteration operations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of occurrence of any accident

previously evaluated.

The ability to remove an RHR loop from operation for up to one

hour per eight-hour period is currently allowed by technical

specification 3.13.9.B.b. By adding a reference to LCO [Limiting

Condition for Operation] 3.13.1.A.4. and adding the requirement to

suspend CORE ALTERATIONS to Action 3.13.9.B.a. to be consistent with

3.13.9.B.b., no change in operating practices or plant configuration

is made. By maintaining the requirement to have an RHR loop in

operation during MODE 6, and by requiring CORE ALTERATIONS to be

suspended if an RHR loop is not back in operation after one hour,

adequate corrective actions are implemented until the RHR loop is

restored to operating status. Therefore, operation of the system is

consistent with current technical specifications and this change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not require a physical alteration of the

plant (no new or different equipment will be installed to implement

this change). The Technical Specifications will continue to require

an RHR loop to be in operation during MODE 6, and will only permit

the loop to be not in operation for up to one hour in an eight-hour

period. The Technical Specifications will continue to require

compliance with these limitations and suspension of CORE ALTERATIONS

if an RHR loop is not in operation for more than one hour. Thus,

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

kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not result in a significant reduction in

a margin of safety because it has no impact on any safety analysis

assumptions. The requirement to have an RHR loop in operation during

MODE 6 is maintained, along with the ability to remove RHR from

operation for up to one hour per eight-hour period. If an RHR loop

is not in service beyond 1 hour per TS 3.13.9.B, CORE ALTERATIONS

will be suspended. Therefore, this change does not involve a

significant reduction in a margin of safety.

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

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

Austin, One

[[Page 66706]]

First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request: November 7, 1996.

Description of amendment request: The proposed amendment would

revise Technical Specification 4.2.9, Service and Instrument Air

System, to add an additional air compressor.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Utilizing the existing piping configuration, both the new and

the existing air compressors are capable of supporting either

portion of the Service and Instrument Air System. The addition of

the fourth air compressor will decrease the probability of an

accident previously evaluated, because capacity is being added to

the system. The consequences of an accident previously evaluated

will not be affected by the addition of a fourth air compressor. The

Service and Instrument Air System performs the non-safety related

function of providing compressed air for service use and moisture

free compressed instrument air for control air demands. The

instrument air portion is designed so that its operation is required

for plant reliability, not plant nuclear safety. Safety-related

equipment supplied by instrument air is designed to fail in its safe

condition upon loss of instrument air or, safety-related equipment

(and nonsafety-related equipment determined to be important to

safety) required to operate subsequent to instrument air failure is

supplied by backup nitrogen accumulators.

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

accident from any accident previously evaluated.

The operation of the equipment in the Service and Instrument Air

System is essentially unchanged. The new air compressor is a similar

design (nonlubricated), providing additional air volume at a quality

comparable to the three existing air compressors. Therefore, the

possibility of an accident of a different kind than any previously

evaluated has not been created.

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

The Technical Specification does not specify a margin of safety

for the operation of the Service and Instrument Air System, other

than specifying that [``Instrument and service] air shall be

supplied by three, nonlubricated air compressors, each rated at 70

scfm [standard cubic feet per minute]. Instrument air shall also

pass through a dryer.'' Addition of a fourth air compressor will

increase the available capacity, thus increasing the margin of

safety. Therefore, adding the statement ``and one, nonlubricated air

compressor rated at 100 scfm'' to Technical Specification 4.2.9.

will not reduce 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: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: John N. Hannon.

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: October 4, 1996.

Description of amendment request: The proposed amendment would

revise the surveillance requirements in Technical Specifications (TSs)

4.1.2.3.1, 4.1.2.4.1, 4.5.2.b, and 4.6.2.1.b and associated Bases. The

subject surveillance requirements are applicable to the charging/high

head safety injection pumps, low head safety injection pump, and the

containment quench spray pumps. The proposed changes would replace the

current specific test acceptance criteria contained in these

surveillance requirements with requirements to verify pump performance

in accordance with the Inservice Testing Program, the Emergency Core

Cooling System Flow Analysis, or the Containment Integrity Safety

Analysis, as applicable. The proposed changes would also make minor

editorial changes in these TSs and make conforming changes in the TS

Index pages.

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 change does not result in a modification to plant equipment

nor does it affect the manner in which the plant is operated. Since

the physical plant equipment and operating practices are not

changed, as noted above, there is no change in the probability of an

accident previously evaluated.

The proposed change will not lower the pump performance

operability criteria for the charging/high head safety injection,

low head safety injection and quench spray pumps, as assumed in the

safety analysis. The required values for developed pump head and

flow will continue to satisfy accident mitigation requirements and

will be maintained and controlled in the Inservice Testing (IST)

Programs(s).

Since the proposed change does not lower the pump's performance

acceptance criteria, as assumed in the safety analysis, the

containment depressurization system will continue to meet its design

basis requirements. The proposed change will not impose additional

challenges to the containment structure in terms of peak pressure.

The calculated offsite dose consequences of a design basis accident

(DBA) will remain unchanged since the one hour release duration and

source term remain unchanged. The ability of the emergency core

cooling system (ECCS) subsystems to provide sufficient emergency

core cooling capability in the event of a loss of coolant accident

(LOCA) remains unchanged. Therefore, peak cladding temperatures

during a LOCA will continue to remain within acceptable limits. The

ability of the ECCS subsystems to provide sufficient long term core

cooling capability in the recirculation mode during the accident

recovery period remains unchanged. The charging pumps, as part of

the boron injection system, will continue to provide sufficient flow

to ensure negative reactivity control during each mode of facility

operation. Future changes to the pump head and flow requirements

will be made under the 10 CFR 50.59 process to ensure that the

system performance requirements continue to be met.

The proposed change to the Bases section will ensure that safety

analyses assumptions for assumed pump performance continue to be

met. The words ``required developed head'' will be clearly defined

to reflect that they refer to the value(s) assumed in the safety

analysis for the pump's developed head at a specific or a given

point. The proposed changes to the Index pages and the footnote in

LCO 3.1.2.4 are administrative in nature and do not affect plant

safety.

Based on the above discussion, it is concluded that this change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not alter the method of operating the

plant. The charging pumps will continue to be in service during

plant operation and be available to perform their function as high

head safety injection pumps. This proposed change does not pose

additional challenges to the design or function of the charging

pumps. The low head safety injection and quench spray

[[Page 66707]]

systems are accident mitigation systems and are normally in standby.

System operation would be initiated as required to mitigate the

consequences of a DBA. The charging/high head safety injection, low

head safety injection and quench pumps will continue to provide

sufficient flow to mitigate the consequences of a DBA. These

systems' operation continues [sic] [continue] to fulfill the safety

functions for which they were designed and no changes to plant

equipment will occur. As a result, an accident which is new or

different than any already evaluated in the Updated Final Safety

Analysis Report will not be created due to this change.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety? The surveillance requirements for demonstrating that the

pumps are operable will continue to assure the ability of the system

to satisfy its design function. Therefore, the proposed change will

not affect the ability of these systems to perform their safety

function.

The containment systems' design requirements to restore the

containment to subatmospheric condition within one hour will

continue to be satisfied. This proposed change does not have an

effect on the containment peak pressure since the charging/high head

safety injection, low head safety injection and quench spray pumps'

performance requirements are not being lowered. The ability of the

ECCS subsystems to provide sufficient emergency core cooling

capability in the event of a LOCA remains unchanged. Therefore, peak

cladding temperatures during a LOCA will continue to remain within

acceptable limits. The ability of the ECCS subsystems to provide

sufficient long term core cooling capability in the recirculation

mode during the accident recovery period remains unchanged. The

charging pumps, as part of the boron injection system, will continue

to provide sufficient flow to ensure negative reactivity control

during each mode of facility operation. There is no resultant change

in dose consequences since source term remains unchanged and the

containment will continue to reach a subatmospheric pressure within

the first hour following a DBA.

Each pump's performance requirements will continue to be

controlled in a manner to ensure safety analysis assumptions are

met.

Therefore, based on the above discussions, it can be concluded

that the proposed change does not involve a significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

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 31, 1996.

Description of amendment request: The proposed amendments will

revise administrative controls Technical Specification (TS) 6.5.1,

``Facility Review Group (FRG),'' and TS 6.8, ``Procedures and

Programs.'' The revisions to TS 6.5.1 reduce the scope of procedures

and procedure changes which require review by the FRG, transfer

approval of certain procedures from the Plant Manager to the FRG, and

require copies of FRG meeting minutes be provided to the Plant Manager.

The changes to TS 6.8 reflect the corresponding changes in TS 6.5.1,

and expand the scope of the section on temporary changes to 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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments revise certain administrative controls

involved with the on-site programmatic process for review and

approval of plant procedures. Specifications that are in place to

provide assurance that the unit operating staff qualifications are

acceptable, and that written procedures are established, implemented

and maintained for safety related activities are not being changed.

The revisions are consistent with industry standards established

pursuant to 10 CFR Part 50, Appendix B, and do not alter any

parameter or equipment performance assumptions that are contained in

plant safety analyses to evaluate the initiation or consequences of

an accident. 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

kind of accident from any accident previously evaluated.

The proposed amendments will not change the physical plant or

the modes of plant operation defined in the Facility License for

either St. Lucie unit. Changes proposed for the administrative

controls do not involve the addition or modification of equipment

nor do they alter the design or operation of plant systems.

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 proposed amendments revise certain administrative controls

involving the on-site programmatic process for review and approval

of plant procedures. The scope, or the requirement to establish,

maintain, and implement procedures for activities that could affect

nuclear safety are not being changed. The proposed changes are

consistent with approved industry standards and do not alter the

basis for any technical specification that is related to the

establishment of, or the maintenance of, a nuclear safety margin.

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

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: 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.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: October 31, 1996 (TSCR 205).

Description of amendment request: The proposed change requests

deletion of Technical Specification Table 3.5.2 which lists automatic

primary containment isolation valves. In addition, this change request

clarifies the applicability of an action statement which applies to

several limiting conditions for operation in Section 3.5 and deletes

closure time requirements for several automatic isolation valves in

Section 4.5.F.

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:

[[Page 66708]]

1. The proposed deletion of the automatic primary containment

isolation valve Table 3.5.2 and closure times for several valves in

Specification 4.5.F.1 are administrative in nature and do not affect

the purpose, function, operability and testing requirements of the

automatic primary containment isolation valves or the isolation

condenser isolation valves. The required action contained in

Specification 3.5.A.7 has been moved to the associated

specifications and has not changed. Capitalizing definitions and

deleting unneeded pages are also administrative changes which

enhance the usability of the Technical Specifications. Therefore,

the proposed changes do not increase the probability of occurrence

or consequence of an accident previously evaluated.

2. The proposed changes are administrative and do not involve a

physical change to plant configuration nor do they affect the

performance of any equipment. Existing limiting conditions for

operation and surveillance requirements are retained. Therefore, the

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

previously evaluated is not created.

3. Deleting the list of valves in Table 3.5.2 and valve closure

times in Specification 4.5.F.1 are administrative changes which do

not affect the purpose or function of the automatic primary

containment isolation valves. The listing of the automatic primary

containment isolation valves and stroke time requirements will be in

controlled plant procedures. Changes to the list or closure times

can be made in accordance with review procedures required by Section

6.5 of the Technical Specifications and 10 CFR 50.59. Similarly,

inserting the statement of required action in Specification 3.5.A.7

into the Specifications to which it applies does not modify the

condition or the action to be taken and is an administrative change

which clarifies the Technical Specifications. Therefore, the margin

of safety is not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: November 12, 1996, as supplemented

November 27, 1996 (TSCR 224).

Description of amendment request: The proposed technical

specification change will reflect the implementation of the revised 10

CFR Part 20, ``Standards for Protection Against Radiation.''

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 of occurrence or the consequences of an accident

previously evaluated.

The proposed revisions to the liquid release rate limits and

bases and gaseous effluent bases will not result in a change in the

types or amounts of effluents released nor will there be an increase

in individual or cumulative radiation exposures. In addition, these

changes do not impact the operation or design of any plant

structures, systems, or components. These changes ensure compliance

with 10 CFR 50.36a and 10 CFR 50 Appendix I and result in levels of

radioactive materials in effluents being maintained ALARA [as low as

is reasonably achievable]. The revision to the high radiation area

controls and dose measurement distance will ensure areas are

conservatively posted as high radiation areas in compliance with 10

CFR 20.1601(a)(1) and provide controls to ensure individuals are not

overexposed. Other proposed changes consist of revisions to 10 CFR

20 references to recognize the new section numbers, and

administrative controls for record keeping to maintain compliance

with the new Part 20.

These changes will not result in a change to plant design or

operation. Therefore, it can be concluded that the proposed changes

do not involve an 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. The

proposed changes do not affect the plant design or operation nor do

they result in a change to the configuration of any equipment. There

will be no change in the types or increase in the amount of

effluents released offsite.

Therefore, this proposed change cannot create the possibility of

a new or different kind of accident from any 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 proposed revisions do not involve any change in the types or

increase in the amount of effluents released offsite. The proposed

changes do not involve any actual change in the methodology used in

the control of radioactive wastes or radiological environmental

monitoring. The methodology that will be used in the control of

radioactive effluents and calculation of effluent monitor setpoints

will result in the same effluent release rate as the current

methodology now being used. The operational flexibility needed for

releases allows the use of limits as proposed. In addition, the

changes in measurement distances for determination of high radiation

areas will not result in an increase in individual or cumulative

occupational radiation exposures since it will result in a more

conservative identification of high radiation areas. Compliance with

the limits of the new 10 CFR 20.1301 will be demonstrated by

operating within the limits of 10 CFR 50 Appendix I and 40 CFR 190.

Thus, operation of the facility in accordance with the proposed

amendment 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: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz.

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

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: August 29, 1996, as supplemented October

3, 1996. The October 3, 1996, submittal contained editorial changes

only and did not change the initial no significant hazards

consideration evaluation.

Description of amendment request: The purpose of this amendment

request is to 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 (SHC), 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 involve a significant increase in the

probability of occurrence or the

[[Page 66709]]

consequences of an accident previously evaluated. The proposed

amendment deletes limiting conditions 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 deletes Surveillance Requirements from the

TMI-1 Technical Specifications that are related to the LCOs to be

deleted. These items are addressed in licensee controlled documents.

Certain design feature specifications are also to be deleted

consistent with the RSTS [Revised Standard Technical Specifications]

for B&W plants. The proposed changes do not modify the operation,

limits or controls of systems, structures or components relied upon

to prevent or mitigate the consequences or 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 because 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 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: 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 & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: John F. Stolz.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: September 20, 1996.

Description of amendment request: The proposed amendment would

revise the Nine Mile Point Unit 1 (NMP1) Technical Specifications that

involve the frequencies of surveillance requirements stated in Tables

4.6.2a, 4.6.2b, 4.6.2g, and 4.6.11, and Sections 4.2.5b(1), 4.3.2b,

4.3.6b(1), 4.3.6b(2), 4.3.6b(3), 4.3.6b(4), 4.3.6c(2), 4.6.13b.1, and

4.6.13b.2. The surveillances associated with these tables and sections

are currently satisfied during NMP1 refueling outages prior to restart

of the unit. The proposed changes would permit surveillance testing

either while the reactor is operating or during outage periods not

associated with refueling. The requirements of the surveillance

sections and tables addressed by this request that are not changed to

be performed at power are being changed to allow surveillance credit to

be taken for performance of the associated surveillances while the

plant is in the Cold Shutdown, Refueling, or Major Maintenance modes.

In addition to these proposed changes, typographical errors are

corrected.

Basis for proposed no significant hazards consideration

determination: The licensee states that: ``The periods between

surveillances will not be inappropriately lengthened. For the affected

surveillances, NMP1 administrative controls will require that the

interval between surveillance testing not exceed a period equal to 1.25

times the nominal 24 months frequency (no longer than 30 months). The

NMP1 plant preventive maintenance and surveillance database will be

revised accordingly.''

The licensee groups the systems affected by this request into four

categories:

Category 1: The associated system will remain operable and able

to automatically perform its safety function during performance of

surveillances that satisfy the proposed surveillance requirement.

Category 2: The system is required for monitoring purposes only

and provides no automatic safety actuation function and redundant,

or redundant and alternate channels are available for required

monitoring.

Category 3: There is no change in the system configuration or

plant operating conditions during the performance of associated

surveillances whether the plant is shutdown for refueling or

shutdown for maintenance. The surveillances performed to meet the

requirements of NMP1 Technical Specifications Tables 4.6.2a

Parameter 8 and 4.6.2g Parameter 6 are included in this category and

may also be completed in concurrence with a unit shutdown. The only

difference between the proposed changes and the normal unit shutdown

sequence is that the mode switch may be taken to ``Shutdown'' in

order to scram the plant. The response of the plant is the same as

it is under the current plant shutdown procedures. There are no

other differences in testing techniques or testing criteria from

those previously required by the NMP1 Technical Specifications.

Category 4: The system or equipment is isolated or out of

service during the performance of the required surveillances. The

associated surveillance may be performed concurrently with quarterly

valve stroking, at which time the system or equipment is already out

of service.

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 operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequence of an accident previously evaluated.

Each of the four categories [* * *] are evaluated separately below:

Category 1: The associated systems will remain operable and able

to automatically fulfill as designed any required safety functions

that may become necessary during performance of required

surveillances. No physical change to the plant design, materials, or

standards is involved. No change to instrumentation operating

characteristics outside current tolerances will be made. No plant

transients will be initiated as a result of the proposed changes. No

initiator of any accident previously evaluated is adversely

affected. No system required to actuate to respond to any accident

previously evaluated in the UFSAR [Updated Final Safety Analysis

Report] is adversely affected by the proposed change.

Category 2: The associated systems will be required for

monitoring purposes only and provide no automatic safety actuation

function and redundant, or redundant and alternate channels are

available for required monitoring. Since redundant monitoring

instrumentation will still be available as required by the technical

specifications, the associated systems' functions in accident

mitigation are not affected. No physical change to the plant design,

materials, or standards is involved. No change to instrumentation

operating characteristics outside current tolerances will be made.

No plant transients will be initiated as a result of the proposed

changes. No initiator of any accident previously evaluated is

adversely affected. No system required to actuate to respond to any

accident previously evaluated in the UFSAR is adversely affected by

the proposed changes.

Category 3: There will be no change in the system configuration

or plant operating conditions during the performance of associated

surveillances. The associated system's ability to perform required

safety functions will not be affected, whether the plant is shutdown

for refueling or shutdown for maintenance. The surveillances

performed to meet the requirements of NMP1 Technical Specifications

Tables 4.6.2a Parameter B and 4.6.2g Parameter 6 are included in

this category and may also be performed in concurrence with a unit

shutdown. The only difference between the proposed changes and the

normal unit shutdown sequence is that the mode switch

[[Page 66710]]

may be taken to ``Shutdown'' in order to scram the plant. The

response of the plant is the same as it is under the current plant

shutdown procedures. There are no other differences in testing

techniques or testing criteria from those previously required by the

NMP1 Technical Specifications. No physical change to the plant

design, materials, or standards is involved. No change to

instrumentation operating characteristics outside current tolerances

will be made. No unexpected plant transients will be initiated as a

result of the proposed changes. No initiator of any accident

previously evaluated is adversely affected. No system required to

actuate to respond to any accident previously evaluated in the UFSAR

is adversely affected by the proposed changes.

Category 4: The associated system or equipment will be isolated

or out of service during the performance of the required

surveillances. The associated surveillances will be performed during

quarterly valve stroking, at which time the system or equipment is

already out of service. No physical change to the plant design,

materials, or standards is involved. No change to instrumentation

operating characteristics outside current tolerances will be made.

No plant transients will be initiated as a result of the proposed

changes. No initiator of any accident previously evaluated is

adversely affected. No system required to actuate to respond to any

accident previously evaluated in the UFSAR is adversely affected by

the proposed changes.

The correction of the typographical errors is administrative only

and has no affect on plant systems or procedures. In all cases,

equipment used for accident mitigation is not adversely affected. The

ability of the operators to safely shut down NMP1 is not impaired. The

changes will not adversely affect any accident precursor or initiator

of any accident. For these reasons, the proposed changes will not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Each of the four categories [* * *] are evaluated separately below.

Category 1: The associated systems will remain operable and able

to automatically perform as designed any required safety functions

that may become necessary during performance of required

surveillances. No physical change to the plant design, materials, or

standards is involved. No change to instrumentation operating

characteristics outside current tolerances will be made. No accident

initiator or failure of a different type than previously identified

in the UFSAR is introduced. No different or new plant transients may

result from those previously evaluated in the UFSAR.

Category 2: The associated systems will be required for

monitoring purposes only and provide no automatic safety actuation

function. Since redundant, or redundant and alternate monitoring

instrumentation will still be available as required by the technical

specifications, the associated systems' functions in accident

mitigation are not affected. No physical change to the plant design,

materials, or standards is involved. No change to instrumentation

operating characteristics outside current tolerances will be made.

No accident initiator or failure of a different type than previously

identified in the UFSAR is introduced. No different or new plant

transients may result from those previously evaluated in the UFSAR.

Category 3: There will be no change in the system configuration

or plant operating conditions during the performance of associated

surveillances. The associated system's ability to perform required

safety functions will not be affected, whether the plant is shutdown

for refueling or shutdown for maintenance. The surveillances

performed to meet the requirements of NMP1 Technical Specifications

Tables 4.6.2a Parameter 8 and 4.6.2g Parameter 6 are included in

this category and may also be performed in concurrence with a unit

shutdown. The only difference between the proposed changes and the

normal unit shutdown sequence is that the mode switch may be taken

to ``Shutdown'' in order to scram the plant. The response of the

plant is the same as it is under the current plant shutdown

procedures. There are no other differences in testing techniques or

testing criteria from those previously required by the NMP1

Technical Specifications. No physical change to the plant design,

materials, or standards is involved. No change to instrumentation

operating characteristics outside current tolerances will be made.

No unexpected plant transients will be initiated as a result of the

proposed changes. No accident initiator or failure of a different

type than previously identified in the UFSAR is introduced. No

different or new plant transients may result from those previously

evaluated in the UFSAR.

Category 4: The associated system or equipment will be isolated

or out of service during the performance of the required

surveillance. The associated surveillances will be performed during

quarterly valve stroking, at which time the system or equipment is

already out of service. No physical change to the plant design,

materials, or standards is involved. No change to instrumentation

operating characteristics outside current tolerances will be made.

No plant transients will be initiated as a result of the proposed

changes. No accident initiator or failure of a different type than

previously identified in the UFSAR is introduced. No different or

new plant transients may result from those previously evaluated in

the UFSAR.

The correction of the typographical errors is administrative only

and has no affect on plant systems or procedures. In all cases, the

changes will not adversely affect any accident precursor or initiator

of any accident and, therefore, the changes do not introduce any new

failure modes or conditions that may create a new or different

accident. For these reasons, the proposed changes do not create the

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

previously evaluated in the UFSAR.

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

Each of the four categories [* * *] are evaluated separately below.

Category 1: The associated systems will remain operable and able

to automatically perform required safety functions during

performance of surveillances that satisfy the surveillance

requirement. There will be no effective change in the interval of

the affected surveillances. The probability of instrument drift or

the ability to detect a failed or drifted instrument remains

unchanged. No physical change to the plant design, materials, or

standards is involved. No change to instrumentation operating

characteristics outside current tolerances will be made. No system

required to actuate to respond to any accident is adversely affected

by the proposed changes. Since each system's operability is not

affected, the margin of safety associated with these systems will

not be significantly reduced.

Category 2: The associated systems will be required for

monitoring purposes only and provide no automatic safety actuation

function. Redundant, or redundant and alternate monitoring

instrumentation will still be available as required by the technical

specifications during the performance of the associated

surveillances. No physical change to the plant design, materials, or

standards is involved. No change to instrumentation operating

characteristics outside current tolerances will be made. There will

be no effective change in the intervals of the affected

surveillances. The probability of instrument drift or the ability to

detect a failed or drifted instrument remains unchanged. No plant

transients will be initiated as a result of the proposed changes. No

initiator of any accident previously evaluated is adversely

affected. No system required to actuate to respond to any accident

is adversely affected by the proposed changes. Therefore, the

associated systems' functions in accident mitigation are not

affected, and no margin of safety will be significantly reduced.

Category 3: There will be no change in the system configuration

or plant operating conditions during the performance of associated

surveillances, the associated system's ability to perform required

safety functions will not be affected, whether the plant is shutdown

for refueling or shutdown for maintenance. The surveillances

performed to meet the requirements of NMP1 Technical Specifications

Tables 4.6.2a Parameter 8 and 4.6.2g Parameter 6 may also be

completed in concurrence with a unit shutdown. The only difference

between the proposed changes and the normal unit shutdown sequence

is that the mode switch may be taken to ``Shutdown'' in order to

[[Page 66711]]

scram the plant. The response of the plant is the same as it is

under the current plant shutdown procedures. There are no other

differences in testing techniques or testing criteria from those

previously required by the NMP1 Technical Specifications. No

physical change to the plant design, materials, or standards is

involved. No change to instrumentation operating characteristics

outside current tolerances will be made. There will be no effective

change in the intervals of the affected surveillances. The

probability of instrument drift or the ability to detect a failed or

drifted instrument remains unchanged. No unexpected plant transients

will be initiated as a result of the proposed changes. No initiator

of any accident if adversely affected. No system required to actuate

to respond to any accident previously evaluated is adversely

affected by the proposed changes. Therefore, no margin of safety

will be significantly reduced.

Category 4: The associated system or equipment will be isolated

or out of service during the performance of the required

surveillances. The associated surveillances will be performed during

quarterly valve stroking, at which time the system or equipment will

already be out of service. No physical change to the plant design,

materials, or standards is involved. No change to instrumentation

operating characteristics outside current tolerances will be made.

There will be no effective change in the intervals of the affected

surveillances. The probability of instrument drift or the ability to

detect a failed or drifted instrument remains unchanged. No plant

transients will be initiated as a result of the proposed changes. No

accident initiator or failure of a different type than identified in

the UFSAR is introduced. Therefore, no margin of safety will be

significantly reduced.

The correction of the typographical errors is administrative only

and has no affect on plant systems or procedures. In all cases, the

changes will not adversely affect any accident precursor or initiator

of any accident and, therefore, the changes do not introduce any new

failure modes or conditions that may create a new or different

accident. None of the proposed changes involve physical modification of

the plant or alterations to any accident or transient analysis.

Therefore, for this and the above reasons, these proposed changes do

not involve any significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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

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

NRC Project Director: S. Singh Bajwa, Acting Director.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook

Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: October 16, 1996.

Description of amendment request: The proposed amendment would

change certain requirements stated in Technical Specification 3/4.8.1,

``AC Sources''. The requirements are related to the emergency diesel

generators (EDGs). The proposed changes would:

1. Increase the EDG fuel storage system minimum volume requirements

specified in Limiting Condition for Operation 3.8.1.1.b.2;

2. Add a footnote applicable to Surveillance Requirement

4.8.1.1.2.f to qualify the words during shutdown. The footnote would

allow the option of performing selected surveillances, or portions

thereof, during conditions or modes other than shutdown;

3. Delete from Surveillance Requirement 4.8.1.1.2.f.14 the

requirement to verify that the cooling tower fans start automatically

on a Tower Actuation signal; and

4. Delete Surveillance Requirement 4.8.1.1.2.h.2 which specifies

performing a periodic pressure test on the ASME Code Class 3 diesel

fuel oil piping.

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.

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

probability or consequences of an accident previously evaluated (10 CFR

50.92(c)(1)).

1. Limiting Condition for Operation 3.8.1.1.b.2

The proposed change increases the minimum EDG fuel oil storage

requirement to account for various factors that may affect the fuel

consumption rate. The revised storage requirement reflects actual EDG

test data and accounts for external variables including fuel oil

specific gravity, heating value of the fuel, and ambient conditions.

The proposed increase in the minimum volume storage requirement is

conservative and ensures that there will be at least a 7 day supply of

fuel oil stored for each EDG to meet the maximum Engineered Safety

Feature load requirements following a loss of power and a design basis

accident as described in Updated Final Safety Analysis Report (UFSAR)

Section 9.5.4.1, Diesel Generator Fuel Oil Storage and Transfer

System--Design Basis. Therefore, the proposed change does not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

2. Surveillance Requirement 4.8.1.1.2.f

The proposed change qualifies the requirement to perform EDG

surveillance requirements ``during shutdown''. Because the terms Hot

Shutdown and Cold Shutdown are defined in the TSs as operating modes or

conditions, the requirement to perform certain surveillances during

shutdown may be misinterpreted, as noted in NRC Generic Letter 91-04.

The proposed footnote would permit certain maintenance and testing

activities to be performed during conditions or modes other than

shutdown. The proposed footnote to Surveillance Requirement 4.8.1.1.2.f

would not alter the intent or the method by which the surveillances are

conducted, and the acceptance criteria for the surveillances would be

unchanged. The footnote would not degrade the ability of the EDGs to

perform their intended function, and it would not affect the response

of the EDGs to a loss of power as described in the UFSAR. Since plant

response to an accident would not change and since failure of an EDG

could not initiate any of the accidents evaluated in the UFSAR, the

proposed footnote would not alter the probability or consequences of an

accident previously analyzed.

3. Surveillance Requirement 4.8.1.1.2.f.14

The cooling tower functions as the ultimate heat sink following a

seismic event which results in blockage of the circulating water

tunnels and therefore a loss of service water. Amendment 18 eliminated

the requirement for automatic start of the cooling tower fans;

therefore, the automatic-start function for the cooling tower fans has

been defeated by placing the control switch in ``Pull-to-Lock''. The

proposed change to delete the automatic fan start reference from

Surveillance Requirement 4.8.1.1.2.f.14 is administrative only to

correct an oversight since the requirement should have been deleted

with the issuance of Amendment 18. The proposed deletion does not

affect the manner by which the facility is operated or involve any

[[Page 66712]]

changes to equipment or features which affect the operational

characteristics of the facility. Since there is no change to the

facility or operating procedures, there is no affect upon the

probability or consequences of any accident previously analyzed.

4. Surveillance Requirement 4.8.1.1.2.h.2

The ASME Code, Section XI, including applicable ASME Code Cases as

authorized by the NRC, provides alternate test methods to use in lieu

of a 110% hydrostatic pressure test that is not practical to perform on

the EDG fuel oil system as currently designed. With the proposed

deletion of Surveillance Requirement 4.8.1.1.2.h.2, the provisions of

Surveillance Requirement 4.0.5 and the ASME Code along with NRC-

authorized Code Cases would be utilized as an equivalent testing

requirement to ensure the continued integrity of the diesel fuel oil

system. Therefore, since the reliability of the EDG fuel oil system

will not be reduced, the probability or consequences of any accident

previously evaluated is not increased.

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

kind of accident from any accident previously evaluated (10 CFR

50.92(c)(2)).

1. Limiting Condition for Operation 3.8.1.1.b.2

The proposed minimum fuel storage requirement has been developed

using actual EDG performance data and accounting for possible

variations in fuel oil specific gravity, heating value of the fuel, and

ambient conditions. The proposed change will provide additional

assurance that there will be at least a 7 day supply of fuel oil to

meet the maximum Engineered Safety Feature load requirements following

a loss of power and a design basis accident. The amount of fuel oil

stored has no effect upon the initiation of any accident sequence,

therefore, the proposed change does not create the possibility of a new

or different kind of accident from any previously analyzed.

2. Surveillance Requirement 4.8.1.1.2.f

The proposed change to allow the option (as supported by a 10 CFR

50.59 safety evaluation) of performing selected surveillance tests, or

portions thereof, during conditions or modes other than during shutdown

does not affect the operation or response of any plant equipment,

including the EDGs, or introduce any new failure mechanism. Therefore,

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

different kind of accident from any previously analyzed.

3. Surveillance Requirement 4.8.1.1.2.f.14

Amendment 18 to the Seabrook Station Operating License approved the

change in the cooling tower operating mode from automatic actuation to

manual actuation. The proposed change to Surveillance Requirement

4.8.1.1.2.f.14 does not create the possibility of a new or different

kind of accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because it does not affect the manner by which the

facility has been operated since Amendment 18 was issued, involve any

changes to equipment or features which affect the operational

characteristics of the facility, or introduce a new failure mode. The

proposed change merely corrects an oversight in that the requirement

should have been deleted when Amendment 18 was issued.

4. Surveillance Requirement 4.8.1.1.2.h.2

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

kind of accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because it does not affect the manner by which the

facility is operated as assumed in the design analysis or Safety

Evaluation, involve any changes to equipment or features which affect

the operational characteristics of the facility, or introduce a new

failure mode. The proposed change merely provides a practical alternate

test method using methods acceptable per Section XI of the ASME Code,

applicable ASME Code Cases as authorized by the NRC, and Regulatory

Guide (RG) 1.137, ``Fuel-Oil Systems at Nuclear Power Plants,''

Revision 1, October 1979. Therefore, the proposed change does not

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

previously analyzed.

C. The changes do not involve a significant reduction in a margin

of safety (10 CFR 50.92(c)(3)).

1. Limiting Condition for Operation 3.8.1.1.b.2

The proposed change does not reduce the ability of the EDGs to

provide sufficient power for at least 7 days to meet the maximum

Engineered Safety Feature load requirements following a loss of power

and a design basis accident as described in UFSAR Section 9.5.4.1.

2. Surveillance Requirement 4.8.1.1.2.f

The proposed change does not reduce the ability of the EDGs to

provide sufficient power to meet the maximum Engineered Safety Feature

load requirements following a loss of power and a design basis accident

as described in the UFSAR. Performing certain surveillances during

conditions or modes other than shutdown (as supported by a 10 CFR 50.59

safety evaluation) does not involve a significant reduction in a margin

of safety (10 CFR 50.92(c)(3)) because it does not affect the manner by

which the facility is operated as assumed in the design analysis or

Safety Evaluation, involve any changes to equipment or features which

affect the operational characteristics of the facility. The proposed

change will continue to ensure the reliability of the EDGs to perform

their intended function.

3. Surveillance Requirement 4.8.1.1.2.f.14

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

kind of accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because it does not affect the manner by which the

facility has operated since Amendment 18 was issued, involve any

changes to equipment or features which affect the operational

characteristics of the facility, or introduce a new failure mode. The

proposed change merely corrects an oversight in that the requirement

should have been deleted when Amendment 18 was issued.

4. Surveillance Requirement 4.8.1.1.2.h.2

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

safety (10 CFR 50.92(c)(3)) because it does not affect the manner by

which the facility is operated or involve any changes to equipment or

features which affect the operational characteristics of the facility.

The proposed change will continue to ensure the reliability of the EDG

fuel oil system. The proposed change merely provides a practical

alternate test method using methods acceptable per Section XI of the

ASME Code, applicable ASME Code Cases as authorized by the NRC, and

Regulatory Guide (RG) 1.137, ``Fuel-Oil Systems at Nuclear Power

Plants,'' Revision 1, October 1979.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities

[[Page 66713]]

Service Company, Post Office Box 270, Hartford CT 06141-0270.

NRC Project Director: S. Singh Bajwa, Acting.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook

Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: October 17, 1996.

Description of amendment request: The proposed amendment would

delete certain instrumentation requirements stated in Technical

Specification (TS) 3/4.3, Instrumentation. The deleted requirements

would be relocated to the Seabrook Station Technical Requirements

Manual (SSTR). The associated Bases for the deleted TS requirements

will be deleted also, but they will not be incorporated into the SSTR.

The following Limiting Conditions for Operation (LCO) and associated

Surveillance Requirements (SRs) would be relocated to the SSTR:

------------------------------------------------------------------------

Technical specification Title

------------------------------------------------------------------------

LCO--3.3.3.2.............................. Incore Detector System.

LCO--3.3.3.3 and associated SRs & Tables.. Seismic Instrumentation.

LCO--3.3.3.4 and associated SRs & Tables.. Meteorological

Instrumentation

LCO--3.3.4 and associated SRs............. Turbine Overspeed

Protection.

------------------------------------------------------------------------

The proposed amendment would also delete (without relocating to the

SSTR) the reference to the location of the meteorological tower from TS

5.5.

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.

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

probability or consequences of an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed changes do not involve any physical

changes to the plant, do not alter the way any structure, system or

component functions, do not modify the manner in which the plant is

operated, do not impact the physical protective boundaries of the

plant, and do not decrease the effectiveness of administrative controls

for assuring safe operation of the facility. The instrumentation-

related systems are not considered a design feature or an operating

restriction that is an initial condition of a design basis accident or

transient analysis, nor do they function in any way to mitigate the

consequences of a design basis accident or transient.

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

kind of accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because the proposed changes do not involve any physical

changes to the plant, do not alter the way any structure, system or

component functions, do not modify the manner in which the plant is

operated, do not impact the physical protective boundaries of the

plant, and do not decrease the effectiveness of administrative controls

for assuring safe operation of the facility.

C. The changes do not involve a significant reduction in a margin

of safety (10 CFR 50.92(c)(3)) because the proposed changes do not

involve any physical changes to the plant, do not alter the way any

structure, system or component functions, do not modify the manner in

which the plant is operated, do not impact the physical protective

boundaries of the plant, and do not decrease the effectiveness of

administrative controls for assuring safe operation of the facility.

Further, the proposed changes do not affect the ability of systems,

structures or components important to safety to perform their intended

function.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.

NRC Project Director: S. Singh Bajwa, Acting.

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: March 20, 1996 and as supplemented on

July 25, 1996.

Description of amendment request: The amendments would modify the

Susquehanna Steam Electric Station (SSES), Units 1 and 2, Technical

Specifications to change the ``open'' logic for the high pressure core

injection (HPCI) suction valves HV-155/255-F042 in order to eliminate

the HPCI pump auto-transfer on high suppression pool level.

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.

Based on the following discussion for the containment, reactor

building, HPCI and RCIC [reactor core isolation cooling] systems,

and the safety-related valves in piping connected to the suppression

pool, the proposed action does not increase the probability or

consequences of an accident previously evaluated. Primary

Containment and Reactor Building Safety-Related Systems, Structures,

and Components Affected by LOCA/SRV [Loss-of-coolant-accident/safety

relief valve] Hydrodynamic Loads

As discussed in the Safety Assessment for this change,

elimination of the HPCI auto suction transfer on high suppression

pool level will allow higher suppression pool water levels in

accidents and transients which involve HPCI operation. The impact of

the higher suppression pool levels were examined for the following

design-basis accidents and transients:

Loss of Coolant Accidents inside containment (FSAR [Final Safety

Analysis Report] *6.2.1.1.3.3),

Inadvertent Safety/Relief valve opening (FSAR *15.1.4),

Primary system break outside containment (FSAR *3.6A),

Inadvertent HPCI initiation (FSAR *15.5.1),

Loss of feedwater flow (FSAR *15.2.7),

Loss of Offsite AC Power (FSAR *15.2.6),

Loss of Main Condenser vacuum (FSAR *15.2.5),

Inadvertent MSIV closure (FSAR *15.2.4),

Turbine trip (with and without bypass) (FSAR *15.2.3),

Generator Load Rejection (with and without bypass), (FSAR

*15.2.2), and

Pressure regulator failure-closed/open (FSAR *15.2.1 & 15.1.3).

These accidents and transients were selected for evaluation

because they involve an initiation of the HPCI system either

inadvertently or as a result of a decrease in vessel inventory and/

or coolant level. Two special events, ATWS and SBO, are also

considered along with the design basis events listed above.

It was concluded that design-basis SRV and LOCA loads envelop

the loads expected with the proposed change. Therefore, the proposed

change does not increase the failure probability of any primary

containment or reactor building structure, system or component which

is affected by LOCA/SRV hydrodynamic loads. The major findings which

lead to this conclusion about SRV and LOCA loads are summarized

below:

DBA [design basis accident] dynamic pressure loads are based on

a maximum initial suppression pool level of 24 feet. The proposed

modification to the HPCI suction

[[Page 66714]]

transfer logic does not affect the initial pool level or the initial

suppression chamber air space volume. During normal plant operation,

suppression pool level (and hence suppression chamber air space

volume) is controlled by Technical Specification requirements.

For LOCAs other than the DBA, the containment is designed for

ADS [automatic depressurization system] blowdown loads in

combination with the LOCA loads. For an intermediate break, the

proposed HPCI modification does allow suppression pool level to

exceed 24 feet by a small amount. ADS loads are, however,

independent of suppression pool level when the downcomer vents are

cleared. Therefore, the proposed modification has no influence on

ADS hydrodynamic loads for an intermediate break.

For small breaks, HPCI injection prevents ADS actuation.

Nevertheless, SRV actuations occur during the RPV [reactor pressure

vessel] cooldown. Downcomer vents are opened in the beginning part

of the accident, but close later on as the break enthalpy decreases.

When the downcomer vents are cleared, the level inside the SRV

tailpipe is not influenced by pool level, and therefore, the SRV

hydrodynamic loads are unaffected by the proposed modification.

During the phase of the accident in which the downcomer vents are

sealed with water, there are no wetwell LOCA hydrodynamic loads, but

the SRV loads are dependent on SP [suppression pool] water level. In

this case, SRV loads are acceptable because SP water level is always

below the Load Limit curve.

ADS actuation would be required in the event of a HPCI failure

during a small-break accident. If HPCI fails during the phase of the

accident in which the downcomer vents are cleared, then ADS loads

would be acceptable because water level (and air volume) within the

SRV tailpipes is independent of pool level. Even if HPCI failure

occurs in the latter part of the accident where the downcomer vents

are sealed, ADS loads are acceptable because water level is always

well below the Load Limit curve.

Under non-LOCA conditions, the containment is designed for

simultaneous actuation of all 16 SRVs. The Load Limit Line defines

the acceptable operating region, in terms of reactor pressure and

suppression pool level, for actuation of all 16 SRVs. Following a

plant transient involving HPCI operation, the suppression pool level

is always below the Load Limit curve, and only a small number of

SRVs actuate to remove decay heat from the reactor.

HPCI System

The proposed change does not increase the probability of an

equipment malfunction in the HPCI system. In fact, the change

eliminates the potential failure of the HPCI suction auto-transfer

on high suppression pool level since that logic is removed.

Potential spurious auto-transfer associated with high suppression

pool logic is also eliminated. HPCI suction auto-transfer on low CST

[condensate storage tank] level and its potential to fail are

unchanged by this change. Also, the change does not affect the

manual suction transfer from the CST to the suppression pool.

As discussed in the safety assessment for this change, the

proposed change has no adverse effects on HPCI valves, pump, or

turbine. Therefore, elimination of the HPCI suction auto transfer

logic (on high suppression pool level) does not increase the

probability of a HPCI malfunction. The consequence of a HPCI failure

in a design-basis accident is evaluated in NEDC-32071P Rev.1,

``Susquehanna Steam Electric Station Units 1 and 2 SAFER/GESTR-LOCA

Loss-of-Coolant Accident Analysis.'' With regard to the fuel, the

consequence of a HPCI failure is unaffected by the proposed change.

If HPCI fails in a design-basis small break accident, ADS

actuation would be required. ADS loads continue to be enveloped by

design loads with the proposed change. Therefore, the proposed

change does not increase the consequences of a HPCI failure.

HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center

1D264(2D264)

On a component level, the failure probability and consequences

of failure associated with the AX [auxiliary] relay in 250 VDC

Control Center 1D264 (2D264) are eliminated because the relay is

disconnected and removed by this modification. Since the control

functions of K19 in panel 1C620 (2C620) have been eliminated, the

failure of the relay has no effect on HPCI suction valve F042

operation.

The 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel

1C620 (2C620) both receive power from battery systems during Station

Blackout. Removal of the relay from 250 VDC Control Center 1D264

(2D264) and the replacement of the relay in HPCI Relay Panel 1C620

(2C620) decreases the load on the battery systems by a small amount.

The change in battery load and line voltage drop is negligible and

is documented in applicable calculations. Dynamic qualification of

the subject equipment is not adversely affected by this modification

as documented in applicable calculations.

RCIC Turbine

As discussed in the safety assessment for this change, RCIC is

used to provide coolant makeup following a reactor vessel isolation

and for an Appendix R shutdown scenario. The Appendix R event also

assumes the reactor vessel is isolated. These events are discussed

in Section 15.2.4 of the FSAR and in the FPRR [fire protection

review report]. The proposed change has no adverse effects on RCIC

turbine operation following a MSIV [main steam isolation valve]

closure (see discussion in the safety assessment for this change

[letter dated March 20, 1996, as supplemented July 25, 1996]).

Therefore, there is no increase in the RCIC failure probability for

the MSIV-closure event or the Appendix R shutdown scenario. The

consequence of RCIC failure is unchanged by the proposed

modification; if RCIC fails, HPCI is available as a backup

system.\1\ [All footnotes are listed at the end of the no signficant

hazards basis section.]

Although RCIC is not designed for mitigation of a small break

accident, the effect of the proposed change on RCIC turbine

operation for such an accident was evaluated in the safety

assessment for this change. The assessment concludes that the

proposed change has no adverse effects on RCIC operation, and

therefore, there is no increase in RCIC failure probability during a

small break accident. Failure of RCIC in a small break accident

would require ADS initiation only for a particular break flow which

is slightly greater than HPCI injection capability. But ADS

initiation has already been considered when evaluating the

consequences of HPCI failure during a small break accident.

Safety-Related Valves on Piping Connected to Suppression Chamber

MOVs [motor operated valves]--The proposed change could

potentially lead to a maximum suppression pool level of 26 feet in a

design-basis accident. This is 2 feet above the maximum design level

of 24 feet. As discussed in the safety assessment for this change,

this is equivalent to a pressure increase of 0.86 psi at the bottom

of the suppression pool. This small pressure increase has negligible

effect on valve operation, and therefore, there is no increase in

the probability of a failure or malfunction of valves in piping

connected to the suppression pool.

Vacuum Breakers--Allowing suppression pool level to potentially

increase to 26 feet in a design-basis accident does not affect the

failure probability of downcomer-vent vacuum breakers because the

level is well below the vacuum breaker elevation of 42 feet.

SRVs/Tailpipes--As discussed in the safety assessment for this

change, the increased suppression pool level associated with the

proposed change does not have any adverse effect on SRV operation or

on the structural integrity of the SRV tailpipe.

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

or different kind of accident from any accident previously

evaluated.

Based on the following discussion for the containment, reactor

building, HPCI and RCIC systems, and the safety-related valves in

piping connected to the suppression pool, the proposed action does

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

from any accident previously evaluated.

The following discussion concerning the impact of the change on

the primary containment, the reactor building, the HPCI system, and

safety-related valves, provides the basis for this conclusion.

Primary Containment and Reactor Building Safety-Related Systems,

Structures, and Components Affected by LOCA/SRV Hydrodynamic Loads

The HPCI suction transfer logic is not necessary to maintain

LOCA loads within design limits because these dynamic pressure loads

are characterized in terms of the SP level at the initiation of the

accident. That is, LOCA blowdown tests were conducted without the

removal of water from the suppression chamber section of the test

tank.2 The increase in pool level realized during these tests

was proto-typical of the pool level increase expected at

Susquehanna. Removal of the HPCI suction transfer logic on high pool

level does not affect suppression pool level at the initiation of a

DBA.3

[[Page 66715]]

In addition, the HPCI suction transfer logic is not necessary to

maintain SRV/ADS blowdown loads within design limits. SRV dynamic

pressure loads consist of two components: air clearing loads and

steam condensation loads. The steam condensation loads are bounded

by the more severe air clearing loads which are caused by gas bubble

oscillations following the expulsion of noncondensible gas from the

SRV tailpipe. Air clearing loads are a function of reactor pressure

and water level inside the SRV tailpipe.

Depending on the break size and location, the downcomer vents

may be cleared for the entire time that HPCI is operating, or they

may reseal in the latter part of the accident. When the downcomer

vents are cleared, the level inside the SRV tailpipe is depressed to

the elevation coinciding with the bottom of the downcomer pipes, and

it is therefore decoupled from the rising suppression pool level. In

this situation SRV air-clearing loads are unaffected by the proposed

change.

When the downcomer vents are sealed with water, the Load Limit

line can be used to determine if SRV/ADS loads are enveloped by

design loads. For the most limiting event, which is the small break

LOCA, the overall safety margin increases as pool level rises during

the event. This is because the decrease in reactor pressure more

than offsets the adverse effects associated with the rise in pool

level.

Since LOCA and SRV dynamic loads remain bounded by design loads,

dynamic loading of primary containment and reactor building

structures, systems, and components are unaffected by the proposed

change. Therefore, with respect to dynamic loads, the proposed

change does not create the possibility for an accident or

malfunction of a different type than any evaluated in the SAR

[Safety Analysis Report].

HPCI System

There are no new HPCI turbine failure modes introduced by the

higher suppression pool levels which can occur with the proposed

change. Turbine exhaust pressure remains well below the design limit

of 65 psia. In addition, the higher pool level does not create the

possibility of water hammer damage to the turbine discharge piping.

If the operator fails to control RPV level less than +54'' (single

operator error) in the long-term part of the small-break accident

when suppression pool level is greater than 25.6 feet, leakage

through check valve F049 is such that it will be contained well

within the volume of the turbine-discharge-line drain pot. Note that

suppression pool level is limited to 26 feet by operator action.

Furthermore, suppression pool level can reach 26 feet only for a

particular range of small breaks, and for this range of small

breaks, suppression pool level would exceed 25.6 feet for only

approximately 10 minutes of the accident duration. This corresponds

to about 10% of the time that HPCI is operating. Thus it is very

unlikely that HPCI would trip with pool level greater than 25.6

feet.

If check valve F049 is failed during the small-break accident

(single equipment failure), the turbine exhaust line would become

flooded if the HPCI system tripped during the 10 minute interval

when suppression pool level greater than 26 feet; however, it is not

necessary to postulate an operator error (failure to control RPV

level less than +54'') along with the check valve failure. A small

break accident with failure of check valve F049 and failure of the

operator to control RPV level as required by the EOPs [emergency

operating procedures], in a narrow time interval during the long-

term part of the accident, is beyond the plant design basis.

A new type of malfunction does not occur even in the beyond-

design-basis condition where failure of check valve F049 is

considered along with failure of the operator to control RPV level

less than 54'' in the narrow time interval when pool level is

greater than 25.6. With these failures, the turbine exhaust piping

will become flooded, and the system may fail on restart. The General

Electric Company has performed an analysis to determine the

consequences of a HPCI start with flooding of the turbine and

adjacent exhaust line.\4\ The analysis, which addresses a potential

design deficiency in the HPCI barometric condenser, shows that the

containment penetration head fitting and interface piping will not

fail as a result of the water hammer associated with the HPCI start.

Since failure of the HPCI system is already considered in the plant

design-basis accident analysis; this is not a different type of

malfunction than that already considered.

HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 1D264(2D264)

No new failure modes are introduced by the hardware changes in

the 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel 1C620

(2C620). Some failure modes are eliminated by the proposed change.

Specifically, the potential failure of the HPCI suction auto-

transfer on high suppression pool level is eliminated since that

logic is removed. Potential spurious auto-transfer associated with

high suppression pool logic is also eliminated. HPCI suction auto-

transfer on low CST level and its potential to fail are unchanged by

this change.

On a component level, potential failure modes for the AX relay

in 250 VDC Control Center 1D264 (2D264) are eliminated by this

modification because the relay is disconnected and removed by this

change. The potential failure modes for the relay K19 in panel 1C620

(2C620) are unchanged. Since the control functions of K19 have been

eliminated, the failure of the relay has no effect on HPCI suction

valve F042 operation.

Removal of the relay from 250 VDC Control Center 1D264 (2D264)

and the replacement of the relay in the HPCI Relay Panel 1C620

(2C620) changes the load on the battery systems by a small amount.

The change in battery load and change in line voltage drop are

negligible and they do not adversely affect the performance of the

panels or battery systems. In addition, seismic qualification of the

panels is not adversely affected by this change.

RCIC Turbine

As discussed in the safety assessment for this change, the

proposed change has no adverse effects on RCIC turbine operation.

Therefore, the proposed change cannot result in a new RCIC failure

mode.

Safety-Related Valves on Piping Connected to Suppression Chamber

MOVs--The increased suppression pool water level which can occur

as a result of the proposed change does not create a failure

mechanism for safety-related valves on piping connected to the

suppression pool. The pressure differential for any valve on piping

connected to the suppression pool will increase by at most 0.86 psi.

This change in differential pressure has negligible effect on valve

operation.

Vacuum Breakers--The proposed change cannot lead to malfunction

of the downcomer-vent vacuum breakers as the maximum level expected

in a design-basis event is 26 feet, and the vacuum breakers are

located at 42 feet above the suppression pool floor.

SRVs/Tailpipes--There is no interaction between increased

suppression pool level and SRV operation since the flow through the

SRVs is choked and therefore decoupled from downstream conditions.

Also, the increased suppression pool level cannot lead to failure of

the SRV tailpipe because the potential level increase is well below

the SRV Tailpipe Level Limit.\5\ If suppression pool water level is

below this limit, there is no concern of tailpipe failure due to

overpressurization. The minimum value of the SRV Tailpipe Level

Limit is 35 feet.\6\ This is 9 feet above the maximum level expected

in a design-basis accident. For beyond-design-basis events, SRV

tailpipe integrity is protected by the EOP requirement to

depressurize the reactor on the SRV Tailpipe Level Limit.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Based on the following discussion for the containment, reactor

building, HPCI and RCIC system, and the safety-related valves in

piping connected to the suppression pool, the proposed action does

not involve a significant reduction in a margin of safety.

HPCI System

The HPCI Technical Specifications ensure that the system is

capable of providing adequate core cooling to limit clad

temperatures in the event of a small break LOCA which does not

result in rapid depressurization of the RPV (Technical Specification

Section 3/4.5.1 & 3/4.5.2). The proposed change has no adverse

affects on the injection capability of the HPCI system. Therefore,

the safety function of the system is not degraded, and there is no

reduction in the margin of safety as defined in the basis for the

HPCI Technical Specifications.

Primary Containment and Reactor Building Safety-Related Systems,

Structures, and Components Affected by LOCA/SRV Hydrodynamic Loads

Removal of the HPCI auto suction transfer on high suppression

pool level does not affect the Technical Specification requirement

to maintain suppression pool water level between 22 and 24 feet

(Technical Specification 3.6.2.1). Therefore, the maximum

containment pressure during the design-basis accident is unaffected

by the proposed change, and there can be no reduction in the margin

of safety as defined in the basis for Technical Specification

[[Page 66716]]

3.6.2.1. Furthermore, a detailed examination of the reactor and

containment response under accident and transient conditions

involving HPCI operation found no situations where the auto suction

transfer was necessary to maintain LOCA and SRV loads within the

design basis envelope. Therefore, from the standpoint of LOCA/SRV

hydrodynamic loads, the proposed change does not reduce the margin

of safety for any primary containment or reactor building structure,

system, or component.

RCIC Turbine

The basis for Technical Specification 3.7.3 states that the RCIC

system is provided to assure adequate core cooling in the event of a

reactor isolation with loss of feedwater flow. The proposed change

does not prohibit RCIC from performing this function, nor does it

degrade in any way the core cooling capability of RCIC. Therefore,

there is no reduction in the margin of safety as defined in the

basis for Technical Specification 3.7.3.

Safety-Related Valves on Piping Connected to Suppression Pool

MOVs--The increase in suppression pool water level which can

occur as a result of the proposed change does not reduce the margin

of safety for safety-related valves on piping connected to the

suppression pool. The pressure differential for any valve on piping

connected to the suppression pool will increase by at most 0.86 psi.

This change in differential pressure has negligible effect on valve

operation.

Vacuum Breakers--The proposed change cannot reduce the margin of

safety as discussed in the basis for Technical Specification 3.6.4

because the maximum level expected in a design-basis event is 26

feet which is well below the downcomer-vent vacuum breaker elevation

of 42 feet.

SRVs/Tailpipes--There is no interaction between increased

suppression pool level and SRV operation since the flow through the

SRVs is choked and therefore decoupled from downstream conditions.

Consequently, there is no reduction in the margin of safety as

defined in the bases for Technical Specifications 3.4.2 (safety

valve function) and 3.5.1.d (ADS function). Also, the increased

suppression pool level does not lead to a reduction in the margin of

safety for the SRV tailpipes because the tailpipes can operate

safely with pool levels up to 35 feet. This is nine feet above the

maximum suppression pool level that can occur in a design-basis

accident with the proposed change. For beyond-design-basis events,

SRV tailpipe integrity is protected by the EOP requirement to

depressurize the reactor on the SRV Tailpipe Level Limit.\7\

HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 1D264(2D264)

As discussed previously, removal of the relay from 250 VDC

Control Center 1D264 (2D264) and the replacement of the relay in the

HPCI Relay Panel 1C620 (2C620) changes the load on the battery

systems by a small amount. The change in battery load and change in

line voltage drop are negligible and therefore they do not reduce

the margin of safety for the panels or battery systems. In addition,

seismic qualification of the panels is not adversely affected by

this change so there is no reduction in the margin of safety for

seismic events.

1. DBD041, Rev. 0, p. 1. [design basis document for RCIC system]

2. SSES DAR [design assessment report for suppression pool

hydrodynamic loads], Section 9.4.1

3. Suppression pool level must be maintained less than 24 feet

in accordance with Technical Specification 3.6.2.1.a.

4. GKR-03-001, ``NRC and Utility Notification of Closeout of GE

PRC92-05, Potential Design Deficiency on HPCI,'' January 6, 1993 [GE

letter to PP&L regarding closure of HPCI design issue].

5. This limit is defined in EO-100/200-103 [emergency operating

procedure]

6. Bechtel Calculations PUP-15598-S2 & PUP-15598-S6, and PLE-

15315 (March 2, 1992)

7. The limit is defined in EO-100/200-103 [emergency operating

procedure]

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 Nos. 50-387 and 50-388

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

Pennsylvania

Date of amendment request: October 7, 1996.

Description of amendment request: The amendments would modify the

Susquehanna Steam Electric Station, Units 1 and 2, Technical

Specifications by revising the trip setpoints and allowable values for

the secondary containment isolation ``Refuel Floor High Exhaust Duct

Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct

Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct

Radiation--High'' monitor in Table 3.3.2-2. The change would enhance

the operational efficiency of plant operations by eliminating

compensatory measures which prevent spurious secondary containment

isolations, and initiation of the standby gas treatment system (SGTS)

and recirculation system during refueling activities. This change would

also allow for the use of the hydrogen water chemistry system during

operation.

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 proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to the trip setpoints and allowable values

to the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,

the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor,

and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor

does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The design basis

for the monitors is to monitor radiation in the unfiltered air from

the Zone III exhaust system to provide signals which isolate the

Zone III of the secondary containment on a high radiation condition,

and to initiate SGTS and the Recirculation system to limit offsite

doses to maintain regulatory requirements.

The original setpoints for these monitors were based upon normal

radiological operating conditions and were set at a value to

preclude spurious design actuations by these monitors during normal

plant operations. However, the monitors are designed to detect

radiation associated with certain postulated accident conditions. As

required by the Technical specifications the monitors are operable

when conditions exist that may result in fuel damage events, and

therefore, will perform their design basis function. Consequently,

an increase to the trip setpoints and allowable values is warranted

since the existing setpoints, which are conservatively based on

normal radiological operating conditions, are not related to the

design basis of the monitors. Therefore, based upon the design basis

of the monitors, an increase to the trip setpoints and allowable

values will not result in a decrease of the safety function of the

monitors but will make the trip setpoints and allowable values

consistent with the design basis.

Based on the design basis of these monitors, revised analytical

limits were derived reflecting the accident function of the

monitors. The analytical limit calculations utilized FSAR realistic

source terms, instead of the worst case source terms utilized for

10CFR [Part] 100 compliance. Use of the realistic source terms

results in conservative analytical limits.

The ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,

and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' [monitor]

are required to be OPERABLE during CORE ALTERATIONS (except for

single control rod movements unless performing TS 3.10.3),

operations with the potential for draining the reactor vessel, and

handling of irradiated fuel in the secondary containment. The

``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor is

required to be operable during handling of irradiated fuel. These

Technical Specification

[[Page 66717]]

applicable operational conditions for the monitors are not affected

since this proposed revision only revises the trip setpoints and

allowable values to be consistent with the design bases of the

monitors.

For the reasons stated above the revisions to the trip setpoints

and allowable values to the ``Refuel Floor High Exhaust Duct

Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct

Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct

Radiation--High'' monitor in Technical Specification.

Table 3.3.2-2 can be implemented without a significant increase

in the probability or consequence of an accident previously

evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident previously evaluated.

The proposed change to the trip setpoints and allowable values

for the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,

the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor,

and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor

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

accident from any accident previously evaluated.

The monitors are designed to limit the release of airborne

radioactivity in the secondary containment Zone III exhaust system

by isolating Zone III, initiating [the] SGTS and initiating the

Recirculation System on high radiation resulting from fuel handling

accidents. Therefore, the design basis for these monitors is to

monitor radiation in the unfiltered air from the Zone III exhaust

system, and provide signals to limit offsite doses to maintain

regulatory requirements. Zone III includes the Refueling Floor and

can include the Railroad Access Shaft during certain alignments.

These radiation monitors are not provided for occupational

protection associated with operational radiation doses. The proposed

revision does not affect the design basis of the monitors nor the

kind of accident associated with the basis; therefore, no potential

to create a new or different accident exists.

For the reasons stated above the revisions to the trip setpoints

and allowable values to the ``Refuel Floor High Exhaust Duct

Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct

Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct

Radiation--High'' monitor in Technical Specification Table 3.3.2-2

can be implemented without creating the possibility of a new or

different kind of accident from any accident previously evaluated.

III. This proposal does not involve a significant reduction on a

margin of safety.

The proposed change to the trip setpoints and allowable values

for the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,

the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor,

and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor

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

The monitors are designed to limit the release of airborne

radioactivity in the secondary containment Zone III exhaust system

by isolating Zone III, initiating [the] SGTS and initiating the

Recirculation System on high radiation resulting from fuel handling

accidents. Therefore, the design basis for these monitors is to

monitor radiation in the unfiltered air from the Zone III exhaust

system, and provide signals to limit offsite doses to maintain

regulatory requirements. Zone III includes the Refueling Floor and

can include the Railroad Access Shaft during certain alignments.

These radiation monitors are not provided for occupational

protection associated with operational radiation doses. However, the

original setpoints for these monitors were conservatively based upon

normal radiological operating conditions and were set at a value to

preclude spurious design actuation by these monitors during normal

plant operations. The calculations performed to support the trip

setpoint and allowable value revisions concluded that the change

will maintain offsite doses within the 10CFR100 limits. The ``Refuel

Floor High Exhaust Duct Radiation--High'' monitor, and the ``Refuel

Floor Wall Exhaust Duct Radiation--High'' are required to be

OPERABLE during CORE ALTERATIONS (except for single control rod

movements unless performing TS 3.10.3), operations with the

potential for draining the reactor vessel, and handling of

irradiated fuel in the secondary containment. The ``Railroad Access

Shaft Exhaust Duct Radiation--High'' monitor is required to be

operable during handling of irradiated fuel. These Technical

Specification applicable operational conditions for the monitors are

not affected since the proposed revision only revises the trip

setpoints and allowable values to be consistent with the design

bases of the monitors.

The proposed revisions to the trip setpoints and allowable

values, in addition to being based on the appropriate accident

conditions, were also developed utilizing standard setpoint change

methodologies that consider instrument and calibration accuracies

and instrument drift tolerances. This provides added conservatism to

assure that the revised trip setpoints and allowable values are not

exceeded.

For the reasons stated above the revisions to the trip setpoints

and allowable values to the ``Refuel Floor High Exhaust Duct

Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct

Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct

Radiation--High'' monitor in Technical Specification Table 3.3.2-2

can be implemented without involving 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.

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

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania.

Date of amendment request: November 25, 1996.

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise the wording in TS Section

4.8.1.1.2.e.2 and the associated TS Bases Section 3/4.8 to remove the

specific reference to the Residual Heat Removal pump motor and its

corresponding kW rating value, and replace it with wording consistent

with that specified in the Improved TS (i.e., NUREG-1433, Revision 1,

``Standard Technical Specifications General Electric Plants,'' dated

April 1995).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. 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 make any physical alterations or

modifications to the plant systems or equipment. The proposed

changes do not adversely impact the operation of any plant

equipment. The EDGs will continue to function as designed to ensure

that the necessary electrical power is provided to essential plant

equipment to mitigate the consequences of an accident, e.g., Loss-

of-Offsite-Power (LOOP) and Loss-of-Coolant Accident LOCA)

coincident with a LOOP (LOCA/LOOP). The proposed TS changes do not

impact the performance testing requirements associated with the

EDGs. The accident mitigating capabilities of the diesel generators

and emergency loads will remain the same.

The proposed TS changes are consistent with the guidance

stipulated in NUREG-1433, Revision [1], ``Standard Technical

Specification General Electric Plants,'' regarding single load

rejection testing of the EDGs. Specifically, the proposed changes

involve revising the wording in TS Surveillance Requirement (SR)

4.8.1.1.2.e.2 to remove the specific reference to the Residual Heat

Removal (RHR) pump motor and associated kW loading value (992 kW),

and replace it with wording indicating that the EDGs must be capable

of rejecting the single largest post-accident load, which is

consistent with NUREG-1433, Revision 1, guidance. The proposed

changes will also provide additional flexibility for future plant

maintenance activities.

Each EDG will continue to be tested by rejecting a load of

greater than or equal to that of its single largest post-accident

load while maintaining voltage and frequency

[[Page 66718]]

within the current specified parameters. The RHR pump motors are

currently used in performing the EDG single load rejection testing.

The RHR pump motors will continued [sic] [continue] to be used in

performing the surveillance testing since they are the single

largest post-accident electrical load. The consequences of a

malfunction of equipment are not affected. Failure of a EDG or its

safety-related loads is bounded by the loss of a Class 1E electrical

power division which has been previously evaluated as discussed in

LGS Updated Final Safety Analysis Report (UFSAR) Sections 8.1.5.2.e

and 8.3.1.1.3.

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 do not make any physical alterations or

modifications to the plant systems or equipment. The proposed

changes do not adversely impact the operation of any plant

equipment. The EDGs will continue to function as designed to provide

essential electrical power to mitigate the consequences of an

accident. The proposed TS changes are consistent with the guidance

stipulated in NUREG-1433, Revision 1, regarding single load

rejection testing of the EDGs. The proposed changes do not introduce

any new accidents or transients. The proposed TS changes will

provide additional flexibility for future maintenance activities.

The proposed changes do not alter any EDG testing requirements or

frequencies. The RHR pump motors are currently used in performing

the EDG single load rejection testing. The RHR pump motors will

continue to be used in performing the surveillance testing since

they are the single largest post-accident electrical load. The

operation of the EDGs and their corresponding safety-related

electrical loads remain unchanged as a result of the proposed TS

changes.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes do not involve any physical changes to

plant systems or equipment. The proposed TS changes are consistent

with the guidance stipulated in NUREG-1433, Revision 1, ``Standard

Technical Specification General Electric Plants,'' regarding single

load rejection testing of the EDGs. The proposed TS changes will

provide additional flexibility for future plant maintenance

activities. The EDGs will continue to function as designed to

provide essential electrical power to mitigate the consequences of

an accident. The operation of the EDGs and their corresponding

safety-related electrical loads remain unchanged as a result of the

proposed TS changes.

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.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: November 15, 1996.

Description of amendments request: The amendments would eliminate

the containment systems Technical Specification 3.6.2.2. ``Spray

Additive System.'' The specification would be replaced with a new

emergency core cooling system Technical Specification 3.5.6 ``ECCS

Recirculation Fluid pH Control System.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change involves replacement of concentrated

NaOH injected via the containment spray system with trisodium

phosphate (TSP) stored in the containment and dissolved in the sump

recirculation solution to maintain acceptable post accident spray/

recirculation solution chemistry. Deletion of the concentrated NaOH

will eliminate a personnel hazard. The pH control system functions

in response to an accident and does not involve or have any effect

on any initiating event for any accident previously evaluated.

Operation under the proposed amendments will continue to ensure that

iodine potentially released post-LOCA [loss-of-coolant accident] is

retained in the sump solution, and resultant offsite and control

room thyroid doses are within the limits of 10 CFR [Part] 100 and 10

CFR [Part] 50, Appendix A, General Design Criterion [GDC] 19,

respectively.

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

or different kind of accident from any accident previously

evaluated. The deleted equipment is isolated from the remaining

equipment by cut-and-capped piping, determinated and/or spared

cables; and interfaces are analyzed to ensure the remaining required

equipment meets applicable original design requirements. The new

equipment (TSP and baskets) is a passive pH control system and is

supported and analyzed to ensure there are no adverse interfaces

(e.g., pipe break, jet impingement, seismic) with existing

equipment, system, or structures.

3. The proposed change does not involve a significant reduction

in a margin of safety. The slight change in recirculation solution

pH maintains adequate protection against chloride and caustic

induced stress corrosion cracking on mechanical systems and

components, and maintains the capability of the solution to retain

iodine. It does not result in a change to the hydrogen generation

analysis for containment. The increased mass inside containment will

have no significant impact on post-accident flood levels,

recirculation solution boron concentration, or peak clad

temperatures. No other operating parameters for systems, structures,

or components assumed to operate in the safety analysis are changed.

The offsite and control room doses meet the limits of 10 CFR [Part]

100 and GDC 19, respectively. Because the trisodium phosphate is

nonvolatile and the baskets are protected with solid covers and are

located slightly above the floor in the containment where access is

strictly controlled, a surveillance interval of once per refueling

outage provides assurance that the TSP will be available.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201.

NRC Project Director: Herbert N. Berkow.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

[[Page 66719]]

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

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

Date of amendment request: June 3, 1996, as supplemented October

23, 1996.

Description of amendment request: The proposed amendment would

clarify a restriction on shutdown margin monitor operability while

changing modes so that it only limits reactivity changes caused by

boron dilution and rod withdrawal.

Date of publication of individual notice in Federal Register: June

20, 1996 (61 FR 31559).

Expiration date of individual notice: July 22, 1996.

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.

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.

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: October 4, 1996 and

supplemented on November 6, 1996.

Brief description of amendments: The amendments add a Mode of

Applicability to Technical Specification 3.2.3.D, Inoperable Rod

Position Indicator Channels.

Date of issuance: November 25, 1996.

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

Amendment Nos.: 176 and 163.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 17, 1996 (61 FR

54240).

The November 6, 1996, submittal provided additional clarifying

information that did not affect the Commission's initial proposed

finding of no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 25, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

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

Date of application for amendments: June 21, 1996.

Brief description of amendments: The amendments revise Technical

Specification (TS) Section 3/4.9.6, ``Manipulator Crane,'' to make the

wording consistent with the TS Bases description and consistent with

the design of the load handling equipment.

Date of issuance: November 25, 1996.

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

30 days.

Amendment Nos.: 156 and 148.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55031) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 25, 1996.

No significant hazards consideration comments received: No.

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

Black Street, Rock Hill, South Carolina 29730.

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

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

Date of application of amendments: September 17, 1996 (TSC 96-01)

as supplemented October 23, 1996.

Brief description of amendments: The amendments lower the maximum

allowable reactor building pressure, lower the actuation setpoint for

actuation of the reactor building spray system, and modify the

associated TS Bases requirements.

Date of Issuance: November 25, 1996.

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

30 days.

Amendment Nos.: 219, 219, 216.

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

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55031). The October 23, 1996, letter provided clarifying information

that did not change the scope of the September 17, 1996, application

and the initial proposed no signficant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 25, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

[[Page 66720]]

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

Mississippi 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: July 31, 1996, as supplemented

by letters of September 5, October 22, and November 15, 20, and 21,

1996, which supersede the application submitted in the letter of May 9,

1996.

Brief description of amendment: The amendment (1) increased the

safety limit minimum critical power ratio (MCPR) for two loop operation

and single loop operation to 1.12 and 1.14, respectively, and (2) added

two General Electric topical reports to the list of documents

describing the analytical methods used to determine the core operating

limits. The changes are to Section 2.1.1, Reactor Core Safety Limits,

and Section 5.6.5, Core Operating Limits Report (COLR), respectively,

of the Technical Specifications. This amendment would go into effect in

Operating Cycle 9, at the end of the current Refueling Outage 8, and

the plant will have a mixed core of Siemens Power Corporation (SPS)

9 x 9-5 and General Electric (GE) GE11 reload fuel. The licensee also

changed the Bases of the Technical Specifications associated with the

above amendment.

Date of issuance: November 21, 1996.

Effective date: November 21, 1996.

Amendment No: 131.

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

Technical Specifications.

Date of initial notice in Federal Register: September 25, 1996.The

October 22, and November 15, 20, and 21, 1996, submittals provide

clarifying information that did not change the initial determination.

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

Safety Evaluation dated November 21, 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.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: April 15, 1996 (TSCR No. 244).

Brief description of amendment: The amendment revises Specification

5.3.1.B to allow the shield plug and the associated lifting hardware to

be moved over irradiated fuel assemblies that are in a dry shielded

canister within the transfer cask in the cask drop protection system.

Date of Issuance: November 7, 1996.

Effective date: November 7, 1996, to be implemented within 30 days

of issuance.

Amendment No.: 187.

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

Technical Specifications.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20849). The Commission's related evaluation of this amendment and final

determination of no significant hazards consideration addressing

comments received on the proposed no significant hazards consideration

determination are contained in a Safety Evaluation dated November 7,

1996.

No significant hazards consideration comments received: Yes.

Local Public Document Room location: Ocean County Library,

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

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

Date of application for amendment: February 22, 1996, and as

supplemented by letters dated July 24, October 4, November 19 and

November 25, 1996.

Brief description of amendment: The amendment changes Clinton Power

Station Technical Specification (TS) 3.3.8.1, ``Loss of Power

Instrumentation,'' and TS 3.8.1, ``AC Sources-Operating,'' by revising

the setpoint for the degraded voltage protection instrumentation and

modifying or deleting other Loss of Power Instrumentation TS

requirements. In addition, changes were also made to the minimum

required diesel generator voltage specified for certain diesel

generator surveillances.

Date of issuance: December 4, 1996.

Effective date: December 4, 1996.

Amendment No.: 110.

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

Technical Specifications.

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

18168).The letters of July 24, October 4, November 19 and November 25,

1996, provided clarifying information and did not represent significant

changes from the original Federal Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 4, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

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

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: October 11, 1996.

Brief description of amendments: These amendments revise Technical

Specification (TS) 3.9.6, ``Refueling Water Level,'' for San Onofre

Nuclear Generating Station (SONGS), Units 2 and 3. The proposed change

is required to restore certain provisions of the SONGS Units 2 and 3

operating practice that were not incorporated during the conversion to

the improved TS (Amendment Nos. 127 and 116, dated February 9, 1996).

Date of issuance: December 3, 1996.

Effective date: December 3, 1996, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2--134; Unit 3--123.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

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

56251) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 3, 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.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: March 24, 1995, as supplemented by

letter dated July 26, 1996.

Brief description of amendment: The amendment revised Technical

Specification (TS) Surveillance Requirement 4.5.1.1.a.1 to base

accumulator operability on actual parameters (i.e., borated water

volume and nitrogen cover-pressure in the tanks) vs. the absence of

alarms.

Date of issuance: November 22, 1996.

Effective date: November 22, 1996, to be implemented within 30 days

of issuance.

Amendment No.: 103.

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications.

[[Page 66721]]

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

18632) The July 26, 1996, letter provided additional clarifying

information and 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 22, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Dated at Rockville, Maryland, this 11th day of December 1996.

For the Nuclear Regulatory Commission.

Steven A. Varga,

Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor

Regulation.

[FR Doc. 96-31944 Filed 12-17-96; 8:45 am]

BILLING CODE 7590-01-P