Biweekly Notice

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 March 3, 1997, through March 14, 1997. The

last biweekly notice was published on March 12, 1997.

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

[[Page 14458]]

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 April 25, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Docketing and

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

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

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

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

inform the Commission by a toll-free telephone call to Western Union at

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

operator should be given Datagram Identification Number N1023 and the

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

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

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: January 15, 1997

Description of amendments request: The proposed change would revise

the values of the minimum and maximum suppression pool water volumes

corresponding to the upper and lower limits of the suppression water

levels specified in TS 3.6.2.1.a.1 such that the implementation of the

administrative controls will no longer be necessary to ensure

compliance with the Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

[[Page 14459]]

1. The proposed amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change revises the values of the minimum and

maximum suppression pool water volume limits. The water inventory of

the suppression chamber is not a precursor of an accident and,

therefore, cannot increase the probability of an accident previously

evaluated. The pressure suppression chamber water pool mitigates the

consequences of loss-of-coolant accidents (LOCAs), transients, and

other events by providing a heat sink for reactor primary system

energy releases. The proposed minimum and maximum pool water volume

values will be consistent with the current suppression pool water

level limits. No changes to setpoints will be made as a result of

the proposed change. The impact of the proposed change to the

minimum and maximum suppression pool volume limits on the

suppression pool temperatures and pressures following a design basis

LOCA, an SRV [Safety Relief Valve] blowdown event, an Anticipated

Transient Without Scram (ATWS) event, an Appendix R fire event, and

a station blackout event has been evaluated and does not cause

accident parameters to exceed acceptable values. In addition, the

impact the proposed change has on the time to reach cold shutdown

when using the alternate RHR [Residual Heat Removal] shutdown

cooling function is negligible.

The potential impact the proposed change to the suppression pool

water volume limits has on SRV line loads, SRV discharge line

reflood height, wetwell pressurization, suppression pool swell

loads, vent thrust loads, and condensation oscillation and chugging

loads was also reviewed. The proposed change to the suppression pool

water volume limits has no adverse impact on any of these

parameters.

The capability of the suppression chamber water pool to perform

its mitigative functions is not affected by the proposed

change. Therefore, the proposed change does not involve a

significant increase in the consequences of an accident previously

evaluated.

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

a new or different kind of accident from any accident previously

evaluated.

The proposed change revises the values of the minimum and

maximum volume of the suppression chamber water pool. The proposed

change will not alter any physical mechanism by which the

suppression chamber water pool volume is maintained between the

minimum and maximum values. The suppression pool water level will

continue to be maintained between -27 and -31 inches. As a result of

the proposed change there are no physical changes to suppression

chamber components or instrumentation. No new mode of operation is

introduced as a result of the proposed change. Analyses have been

performed which conclude that the proposed change would not affect

the operability of equipment designed to mitigate the consequences

of an accident. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed license amendment[s do] not involve a

significant reduction in a margin of safety.

The proposed change revises the values of the minimum and

maximum suppression chamber water pool volumes. The pressure

suppression chamber water pool mitigates the consequences of several

postulated accidents and transients by providing a heat sink for the

primary coolant system. These accidents and events are the

postulated design basis LOCA, Safety Relief Valve blowdown, ATWS,

Appendix R fire and station blackout events. The consequences of the

proposed change in the suppression pool water volume limits have

been evaluated for these events.

The results of the analyses for the postulated accidents and

events indicate the temperature of the suppression pool water could

increase slightly as a consequence of the decrease in the minimum

suppression pool water volume limit. However, the containment

temperatures remain within acceptable values. The impact of the

calculated increase in containment temperature on the available Net

Positive Suction head (NPSH) for the Residual Heat Removal (RHR) and

Core Spray pumps has been evaluated for the postulated design basis

LOCA and indicate adequate NPSH is maintained throughout the event.

The potential impact of the proposed change to the suppression

pool water volume limits on SRV line loads, SRV discharge line

reflood height, wetwell pressurization, suppression pool swell

loads, vent thrust loads, and condensation oscillation and chugging

loads was evaluated with the conclusion that there are no adverse

impacts on these parameters.

In addition, a small suppression pool water temperature increase

could result due to the reduction in the minimum suppression pool

volume limit in the event reactor shutdown is conducted through a

path utilizing the suppression pool. Such a shutdown path is an

alternative to the normal RHR shutdown cooling function, and the

small potential increase in temperature results in a negligible

increase in the time required to reach cold shutdown conditions.

Cold shutdown conditions could still be reached well within the

Technical Specification requirements.

The proposed increase in the suppression pool water volume limit

does not adversely impact containment parameters as a result of

postulated accidents and events. The potential increase in

temperature of the pressure suppression pool water does not

significantly decrease the ability to maintain containment

parameters within acceptable limits. The potential increase in time

to reach cold shutdown conditions utilizing the suppression pool as

an alternative to the normal RHR shutdown cooling function is

negligible. Therefore, the proposed change to revise the minimum and

maximum suppression water pool volumes 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart (Acting)

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

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

North Carolina

Date of amendment request: March 14, 1997

Description of amendment request: The proposed change revises

Technical Specification 3/4.5.4, ``Refueling Water Storage Tank,'' and

its associated Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The non-safety, non-seismic hydrotest pump is normally

maintained separated from the RWST [Refueling Water Storage Tank] by

a safety-related, locked closed manual operated boundary isolation

valve (1CT-22). However, performance of Technical Specification

required surveillance test OST-1506, ``Reactor Coolant System

Isolation Valve Leak Test - 18 Month Interval- Mode 3,'' requires

the short term use of the hydrotest pump during plant operating

modes. Specifically, this hydrotest pump provides a high pressure

source for leak testing the RCS [Reactor Coolant System] pressure

isolation valves in Mode 3. The test is performed prior to entry

into Mode 2, each refueling outage, whenever flow is established

through the pressure isolation valves, or whenever the plant has

been in cold shutdown for greater than 72 hours. Normally, the test

is completed in less than 8 hours. Due to the piping configuration,

a break in the non-seismic portion of the piping during these

planned evolutions could result in draining the RWST below the

minimum analyzed volume. Therefore to mitigate the consequences of a

failure in the non-seismic piping, manual actions will be needed to

isolate the break flow, (i.e., close valve 1CT-22), prior to

reducing the water volume in the RWST below the minimum analyzed

volume.

[[Page 14460]]

Based on the use of a dedicated attendant to close valve 1CT-22,

the lack of significant accessibility concerns, and the reliability

of the valve to function, it can be concluded that 30 minutes is

ample time for a valve attendant stationed at the valve to execute

the manual action. Since the RWST volume margin provides up to 103

minutes to respond to the pipe failure, it is reasonable to assume

that manual actions to isolate the postulated pipe failure can be

taken before the RWST level decreases below the minimum analyzed

volume assumed in the safety analysis.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Based on the use of a dedicated attendant to close valve 1CT-22,

the lack of significant accessibility concerns, and the reliability

of the valve to function, it can be concluded that 30 minutes is

ample time for a valve attendant stationed at the valve to execute

the manual action. Since the RWST volume margin provides up to 103

minutes to respond to the pipe failure, it is reasonable to assume

that manual actions to isolate the postulated pipe failure can be

taken before the RWST level decreases below the minimum analyzed

volume assumed in the safety analysis. As a result, the capability

of the RWST to perform its safety function is not impacted.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

As described in the Technical Specification Bases, the

operability of the RWST ensures that a sufficient supply of borated

water is available for injection into the core by the emergency core

cooling system. This borated water is used as cooling water for the

core in the event of a LOCA [loss-of-coolant accident] and provides

negative reactivty to counteract any positive increase in reactivity

caused by reactor coolant system (RCS) cooldown. The limits on RWST

minimum volume and boron concentration assure that: (1) sufficient

water is available within containment to permit recirculation

cooling flow to the core, and (2) the reactor will remain

subcritical in the cold condition following mixing of the RWST and

the RCS water volumes with all shutdown and control rods inserted

except for the most reactive control assembly. These limits are

consistent with the assumptions of the LOCA and steam line break

analyses.

Based on the use of a dedicated attendant to close valve 1CT-22,

the lack of significant accessibility concerns, and the reliability

of the valve to function, it can be concluded that 30 minutes is

ample time for a valve attendant stationed at the valve to execute

the manual action. Since the RWST volume margin provides up to 103

minutes to respond to the pipe failure, it is reasonable to assume

that manual actions to isolate the postulated pipe failure can be

taken before the RWST level decreases below the minimum analyzed

volume assumed in the safety analysis. As a result, the capability

of the RWST to perform its safety function is not impacted.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart, Acting

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, Rock Island County,

Illinois

Date of amendment request: February 17, 1997

Description of amendment request: The proposed amendment would

change the required diesel generator load during the initial 2 hours of

a surveillance run from 2625 kW and 2750 kW to 2730 kW and 2860 kW.

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 represent a correction to the emergency

diesel generator surveillance requirement. The proposed changes are

administrative in nature and do not significantly increase the

probability or consequences of any previously evaluated accidents

for Quad Cities Station. The proposed amendment is consistent with

the current safety analyses and represents sufficient requirements

for the assurance and reliability of equipment assumed to operate in

the safety analysis. As such, these changes will not significantly

increase the probability or consequences of a previously evaluated

accident.

The associated systems related to this proposed amendment are

not assumed in any safety analysis to initiate any accident sequence

for Quad Cities Station; therefore, the probability of any accident

previously evaluated is not increased by the proposed amendment.

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

from any accident previously evaluated because:

The proposed amendment for Quad Cities Station's Technical

Specification is required to ensure the diesel generator is tested

in accordance with the design basis requirements. The proposed

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

accident previously evaluated for Quad Cities Station. No new modes

of operation are introduced by the proposed changes. The proposed

changes are administrative in nature and maintain at least the

present level of operability. Therefore, the proposed changes do not

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

any previously evaluated.

The associated systems related to this proposed amendment are

not assumed in any safety analysis to initiate any accident sequence

for Quad Cities Station; 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 is required to ensure the diesel

generator is tested in accordance with the design basis

requirements. The proposed changes are administrative in nature and

do not adversely affect existing plant safety margins or the

reliability of the equipment assumed to operate in the safety

analysis. The proposed changes have been evaluated and found to be

acceptable for use at Quad Cities based on system design, safety

analysis requirements and operational performance. Since the

proposed changes are administrative in nature and maintain necessary

levels of system or component reliability, the proposed changes do

not involve a significant reduction in the margin of safety.

The proposed amendment for Quad Cities Station will not reduce

the availability of systems required to mitigate accident

conditions; therefore, the proposed changes do not involve a

significant reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: 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

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of amendment request: December 24, 1996 and January 31, 1997

[[Page 14461]]

Description of amendment request: Changes to Administrative

Controls section of the Technical Specifications needed to implement

revised management responsibilities and titles that reflect the

permanently shut down status of plant.

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:

In accordance with 10 CFR 50.92, CYAPCO [Connecticut Yankee

Atomic Power Company] and NNECO [Northeast Nuclear Energy Company]

have reviewed the attached proposed changes and have concluded that

they do not involve a Significant Hazard consideration (SHC). The

basis of this conclusion is that the three criterion of 10 CFR 50.92

are not compromised. The proposed changes do not involve an SHC

because the proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No design basis accidents are affected by these proposed

changes. The proposed changes are administrative in nature and are

being proposed to reflect the organizational changes which became

effective December 9, 1996.

The Haddam Neck unit changes are replacement of the Executive

Vice President, Nuclear by the Executive Vice President and Chief

Nuclear Officer along with the replacement of the Vice President,

Haddam Neck by the Unit Director.

No safety systems are adversely affected by the proposed

changes, and no failure modes are associated with the changes.

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

from any accident previously evaluated.

There are no changes in any way that the plants are operated due

to this administrative change. The potential for an unanalyzed

accident is not created. There is no impact on plant response, and

no new failure modes are introduced. The proposed administrative and

editorial changes have no impact on safety limits or design basis

accidents, and have no potential to create a new or unanalyzed

event.

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

These changes do not directly affect any protective boundaries

nor do they impact the safety limits for the protective boundaries.

These proposed changes are administrative and editorial in nature.

Therefore there can be no reduction in the margin of safety.

The Commission has provided guidance concerning the application

of the standards in 10 CFR 50.92 by providing certain examples (51

FR 7751, March 4, 1986) of amendments that are considered not likely

to involve an SHC. The changes proposed herein are enveloped by

example (1), since they are purely administrative changes to the

technical specifications to reflect organizational title changes and

to achieve consistence throughout the technical specifications at

Haddam Neck.

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: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270

NRC Project Director: Seymour H. Weiss

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

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

Date of amendment request: July 17, 1996

Description of amendment request: The proposed change request

modifies Waterford Steam Electric Station, Unit 3, Technical

Specifications 3/4.7.1.3,'' CONDENSATE STORAGE POOL,'' by increasing

the minimum required contained water volume from 82 percent to 91

percent indicated level. This proposed change is required to ensure

that the minimum useable water volume in the Condensate Storage Pool

(CSP) is maintained greater than or equal to 170,000 gallons. The new

minimum level accounts for the minimum level required to prevent

Emergency Feedwater pump suction line vortexing and instrument

measurement uncertainties.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

Increasing the minimum required CSP level will insure that the

minimum required 170,000 gallons of water is available for supply to

the Emergency Feedwater System. Maintaining the minimum required

water volume will not increase the probability of any accident

previously evaluated. Additionally, it will not affect the

consequences of any accident. Maintaining at least 170,000 gallons

of water available in the CSP will ensure that the system remains

within the bounds of the accident analysis. Therefore, the proposed

change will not involve a significant increase in the probability or

consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different type of

accident from any accident previously evaluated?

Response: No.

Increasing the minimum water volume of the CSP from 82 percent

to 91 percent does not create a possibility for a new or different

kind of accident. The CSP will be operated in the same manner as

previously evaluated. Therefore, the proposed change will not create

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

accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

Operation in accordance with this proposed change will ensure

that the minimum contained water volume of the CSP will remain at

least 170,000 gallons under all conditions. This will maintain the

present margin of safety. Therefore, the proposed change will not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: February 5, 1997

Description of amendment request: The proposed amendment will

change Waterford Steam Electric Station, Unit 3, Technical

Specifications 3.1.2.7, 3.1.2.8, 3.5.1, 3.5.4, 3.9.1, and Bases 3/

4.1.2. The proposed change will increase the minimum boron

concentration in the Safety Injection Tanks (SITs) and the Refueling

Water Storage Pool (RWSP) to 2050 ppm to reflect the safety analysis

for fuel Cycle 9.

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 14462]]

The Safety Injection System (SIS) is designed to provide core

cooling in the unlikely event of a loss of coolant accident (LOCA).

The cooling must be sufficient to prevent significant alteration of

core geometry, preclude fuel melting, limit the cladding metal-water

reaction, and remove the energy generated in the core for an

extended period of time following a LOCA. The SIS fluid must contain

the necessary boron concentration to maintain the core subcritical

for the duration of a LOCA.

The proposed change increases the minimum boron concentration in

the SITs and RWSP from 1720 ppm to 2050 ppm. Thus, the SIT/RWSP will

at all times contain sufficient borated water to provide adequate

shutdown margin. Sampling of the system and RWSP required by the

Technical Specifications assures that the required dissolved boron

concentration is present. In addition to its emergency core cooling

function, the SIS functions to inject borated water into the RCS to

increase shutdown margin following a rapid cooldown of the RCS as a

result of a steam line rupture.

Operation of the safety injection system is credited in the

steam line break analysis for causing a decrease in core reactivity.

The current minimum RWSP/SIT concentration to be injected is 1720

ppm. Thus an increase to 2050 ppm will have no adverse affect on

this analysis.

The Mode 5 boron dilution event identifies that with an initial

boron concentration of 1240 ppm, a Keff of 0.98, RCS partially

drained, and one charging pump operational, the minimum possible

time to criticality is greater than 90 minutes. For all other

combinations of Keff, RCS conditions, and number of charging pumps,

the time to loss of shutdown margin is greater than 55 minutes.

Thus, the proposed increase in boron concentration will not affect

the results of the Mode 5 boron dilution event.

The change to the action statement of TS 3.9.1 assures that the

more limiting reactivity condition of a Keff less than 0.95 or a

boron concentration of 2050 ppm specified in the COLR [Core

Operating Limit Report] will be adhered to during refueling

operations.

The upper limit on boron concentration has not changed;

therefore, there will be no affect on boric acid precipitation post-

LOCA.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed change does not physically alter the configuration

of the plant and, therefore, does not create the possibility of a

new or different kind of accident from any previously evaluated

accident.

Therefore, the proposed change will not create the possibility

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

evaluated.

The proposed change maintains the minimum of 55 minutes to

criticality for the refueling mode boron dilution event analysis.

The proposed change continues to ensure that borated water of

sufficient concentration is injected from both the SITs and the RWSP

in the event of a LOCA or MSLB [main steam line break] and that

boric acid does not precipitate in the core during long term cooling

following a LOCA.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: February 10, 1997

Description of amendment request: The proposed amendment would

provide the requirements for avoidance and protection from thermal

hydraulic instabilities as described in NRC Generic Letter 94-02,

``Long-Term Solutions and Upgrade of Interim Operating Recommendations

for Thermal Hydraulic Instabilities in Boiling Water Reactors.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. In fact, it does not result in an increase in the

probability or consequences of any previously evaluated accidents.

The implementation of [Boiling Water Reactor Owners' Group] BWROG

Long-Term Stability Solution Option I-D at [Cooper Nuclear Station]

CNS does not modify the assumptions contained in the existing

accident analysis. The use of an exclusion region and the operator

actions required to avoid and minimize operation inside the region

do not increase the possibility of an accident.

Conditions of operation outside of the exclusion region are

within the analytical envelope of the existing safety analysis. The

operator action requirement to exit the exclusion region upon entry

minimizes the possibility of an oscillation occurring. The actions

to drive control rods and/or to increase recirculation flow to exit

the region are maneuvers within the envelope of normal plant

evolutions. The flow-biased scram has been analyzed and will provide

automatic fuel protection in the event of an instability. Thus, each

proposed Technical Specification requirement provides defense for

protection from an instability event within the existing assumptions

of the accident analysis.

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

new or different kind of accident from any accident previously

evaluated.

As stated above, the proposed Technical Specification

requirements either mandate operation within the envelope of

existing plant operating conditions or force specific operating

maneuvers within those carried out in normal operation. Since

operation of the plant with all of the proposed requirements is

within the existing operating basis, an unanalyzed accident will not

be created through implementation of the proposed change.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

Each of the proposed requirements for plant thermal-hydraulic

stability provides a means for fuel protection. The combination of

avoiding possible unstable conditions and the automatic flow-biased

reactor scram provides an in-depth means for fuel protection.

Therefore, the individual or combination of means to avoid and

suppress an instability supplements 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: William D. Beckner

Northeast Nuclear Energy Company, et al., Docket No. 50-245,

Millstone Nuclear Power Station, Unit No. 1, New London,

Connecticut

Date of amendment request: March 6, 1997

Description of amendment request: During a self assessment, the

licensee identified weaknesses in the current Technical Specifications

regarding allowed outage times for certain specific protective

instrumentation and also for reactor building access control. The

proposed amendment is designed to eliminate these weaknesses by

adopting guidance from NUREG-0123, ``Standard Technical Specifications

for General Electric Boiling Water Reactors (BWR/

[[Page 14463]]

5),'' Revision 3, and NUREG-1433, Standard Technical Specifications

General Electric Plants BWR/4,'' Revision 1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The operation of Millstone Nuclear Power Station, Unit No. 1,

in accordance with the proposed amendment, will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The inherent redundancy and reliability of the protective

instrumentation trip systems ensure that the consequences of an

accident are not significantly increased. In addition, the

restrictive Allowable Outage Time (AOT) interval limits the

probability of the protective instrument channel being unavailable

and an accident requiring its function from occurring

simultaneously. The requirement that the associated trip function

maintains trip capability ensures that the protective

instrumentation response will occur such that the consequences of an

accident are not different from those previously evaluated.

Instruments addressed in the proposed TS respond to changes in

the plant. The proposed (AOTs) provide a two-hour interval where the

instrument is inoperable, yet the Technical Specification (TS)

Limiting Condition for Operation (LCO) action statement is not

immediately entered. The probability of a plant transient being

initiated by a trip of a coincident channel during surveillance

testing is reduced since the channel under test will only be tripped

for a small portion of the test interval. Therefore, AOTs provided

by the proposed TS have no effect on the probability of occurrence

of previously evaluated accidents.

The proposed TS changes provide a two-hour interval where the

instrument is inoperable, but the TS LCO action statement is not

immediately entered. If a single failure occurred on the other

channel of the trip system being tested and the channel being tested

was not in the trip condition, a valid signal might not provide the

required protective action. The probability of an event requiring

initiation of the protective function within the proposed AOT is

low. Additionally, surveillance testing is not generally performed

on multiple sensors simultaneously. So, other trip functions and

sensors remain operable and the probability of extensive

inoperabilities affecting diverse trip functions is low. A spurious

trip of a coincident channel could initiate a plant transient (for

example, a reactor scram or a main steam isolation valve closure);

however, these transients are bounded by the current analyses.

Moreover, the original TS bases submitted as part of the application

for Millstone Unit No. 1's Provisional Operating License (dated

October 7, 1970) included recognition that instruments would be

inoperable during required functional tests and calibrations. Thus,

these conditions were recognized in the original design bases and

constitute part of the licensing bases of the plant. NUREG-0123

provided specific time frames[,] ...AOTs addressed in the table

notes[,] and specific action statements. Millstone Unit No. 1 AOT

values chosen are consistent with these values and less than those

approved in NUREG-1433 which had a more detailed study performed to

lengthen the AOT value.

The existing TS definition for Instrument Functional Test would

be difficult to satisfy if the LCO condition of tripping the

inoperable channel was performed. A similar problem of complying

with the Instrument Calibration definition also exists. The TS

requirement to perform functional tests and calibrations is not

consistent with a requirement to trip the system under test. The

proposed TS changes permit more complete functional and calibration

testing. For example, the main scram contactors could be included

within the surveillance tests. Therefore, these TS clarifications do

not increase the consequences of any previously analyzed accidents.

The two-hour instrumentation AOT for the Air Ejector Off-Gas

System radiation monitors is slightly less restrictive than that

allowed by the NUREG-0123. Since this requirement was relocated from

NUREG-1433, there is no corresponding requirement for comparison.

These radiation monitors are arranged in a two-out-of-two logic;

therefore, both must trip to initiate the required action (closure

of the off-gas isolation valve to the main stack). This action,

however, is automatically delayed by 15 minutes. A high radiation

condition sensed by the monitor in service would provide sufficient

time to take corrective actions. Since a two-hour AOT is deemed

acceptable for instrumentation in system[s] such as the Reactor

Protection System and Emergency Core Cooling Systems, it is

appropriate to apply a two-hour AOT to these radiation monitors.

Additionally, the NUREG-0123 AOT of one hour does not allow

sufficient time to perform required surveillance testing without

placing undue stress on the test performer. The probability of a

plant transient (e.g., loss of condenser vacuum) resulting from a

trip of the coincident channel during surveillance testing is

reduced since the channel under test will only be tripped for a

small portion of the test interval. This transient is bounded by

existing analyses. Therefore, this proposed AOT will not

significantly increase the probability or consequences of an

accident previously evaluated.

Since no physical change is being made to the secondary

containment, or to any systems or components that interface with the

secondary containment, there is no change in the probability of any

accident analyzed in the UFSAR [Updated Final Safety Analysis

Report].

The proposed change continues to ensure the secondary

containment requirements meet the licensing basis. Also, the

proposed changes are based on Standard Technical Specifications,

NUREG-1433, ``Standard Technical Specifications General Electric

Plants, BWR/4,'' Revision 1 guidelines and implement actions to be

taken when secondary containment integrity is not met. If secondary

containment integrity is not met, existing TS 3.7.C directs the

plant to be placed in an operating condition where secondary

containment is not required, e.g., COLD SHUTDOWN. A four hour

allowable outage time is proposed which provides a period of time to

correct the problem that is commensurate with the importance of

maintaining secondary containment during RUN, STARTUP/HOT STANDBY or

HOT SHUTDOWN. The secondary containment is not an initiator for any

accident. Therefore, the proposed change will not increase the

probability of any previously analyzed accident. This short time

period ensures that the probability of an accident requiring

secondary containment integrity operability occurring during periods

when secondary containment integrity is inoperable is minimal.

The proposed surveillance requirement is based on the NUREG-1433

surveillance requiring periodic confirmation that at least one door

in each of the double-doored accesses to the secondary containment

is closed, provides additional assurance of secondary containment

system integrity. While this is a deviation from NUREG-1433 (which

requires that both doors in each access be closed except for normal

entry and exit), it is consistent with the current definition of

Secondary Containment Integrity, which requires that at least one

door in each access opening be closed. Hence, the deviation is

justifiable and represents increased passive testing which will

provide increased awareness of plant conditions. Increased awareness

of plant conditions should reduce the probability or consequences of

any accident previously evaluated.

Since the aspects of secondary containment integrity affected by

reactor building access control are being revised in this proposed

amendment to agree with the allowable outage time allowed by NUREG-

1433 upon loss of secondary containment integrity, the change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Since the editorial items do not alter the meaning or intent of

any requirements, they do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The operation of Millstone Nuclear Power Station, Unit No. 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.

The proposed change to the protective instrumentation trip

system specifications do not create the possibility of a new or

different kind of accident because they do not introduce any new

operational modes or physical modifications to the plant.

Instruments addressed in the proposed TS respond to changes in

the plant. The proposed AOTs provide a two-hour interval where the

instrument is inoperable, yet the TS LCO action statement is not

immediately entered. Given a single failure, this could impact the

response of the trip channel but not the initiation of the event.

The only

[[Page 14464]]

action resulting from the AOTs is to perform testing as required by

TS. Spurious signals during testing could initiate transients but

would be bounded by the previous transient analyses. These tests do

not subject the instruments to any conditions beyond their design

specifications and are performed in accordance with approved testing

standards. This testing ensures equipment operability by identifying

degraded conditions, initiating corrective action and properly

retesting them. Therefore, the proposed TS changes will not

introduce a new or different kind of accident than previously

evaluated.

The two-hour instrumentation AOT for the Air Ejector Off-Gas

System radiation monitors is slightly less restrictive than that

allowed by the NUREG-0123. Since this requirement was relocated from

NUREG-1433, there is no corresponding requirement for comparison.

These radiation monitors are arranged in a two-out-of-two logic;

therefore, both must trip to initiate the required action (closure

of the off-gas isolation valve to the main stack). This action,

however, is automatically delayed by 15 minutes. A high radiation

condition sensed by the monitor in service would provide sufficient

time to take corrective actions. Since a two-hour AOT is deemed

acceptable for instrumentation in system[s] such as the Reactor

Protection System and Emergency Core Cooling Systems, it is

appropriate to apply a two-hour AOT to these radiation monitors.

The proposed changes to Millstone Unit No. 1 Technical

Specifications Section 3.7/4.7 and associated bases were developed

using the guidance provided in the Standard Technical

Specifications, NUREG-1433, ``Standard Technical Specifications

General Electric Plants, BWR/4,'' Revision 1. Augmentation of the

existing surveillance requirements by incorporation of an additional

NUREG-1433 based surveillance, provides additional assurance of

secondary containment system integrity. While this is a deviation

from NUREG-1433 (which requires that both doors in each access be

closed except for normal entry and exit), it is consistent with the

current definition of Secondary Containment Integrity which requires

that at least one door in each access opening be closed. Hence, the

deviation is justifiable and represents increased passive testing

which will provide increased awareness of plant conditions.

Increased awareness of plant conditions will not create the

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

previously evaluated. Since the proposed changes do not

significantly degrade the present level of system operability and

add provisions from NUREG-1433, the proposed amendment does not

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

any accident previously evaluated.

Since the editorial items do not alter plant configurations or

operating modes, they do not create the possibility of a new or

different kind of accident.

3. The operation of Millstone Nuclear Power Station, Unit No. 1,

in accordance with the proposed amendment, will not involve a

significant reduction in a margin of safety.

The protective instrumentation surveillance requirements provide

verification of the operability of the trip system instrumentation

channels. In addition, the channel that monitors the identical Trip

Function within the same trip system maintains trip capability for

the relatively short duration that the coincidence change is in

effect. This ensures that protective instrumentation reliability is

maintained. The proposed change provides for a specific time period

to perform required surveillances on instrument channels without

trips present in associated trip systems. This time allotment tends

to enhance the margin of safety by decreasing the probability of

unnecessary challenges to safety systems and inadvertent plant

transients.

The proposed TS provide a two-hour interval where the instrument

is inoperable, yet the TS LCO action statement is not immediately

entered. If a single failure occurred on the other channel of the

trip system being tested and the channel being tested was not in the

tripped condition, a valid signal might not provide the required

protective action. The probability of an event requiring initiation

of the protective function within the proposed AOT is low.

Additionally, surveillance testing is not generally performed on

multiple sensors simultaneously. So, other trip functions and

sensors remain operable and the probability of extensive

inoperabilities affecting diverse trip functions is low.

The existing TS definition for Instrument Functional Test would

be difficult to satisfy if the LCO condition of tripping the

inoperable channel was performed. A similar problem of complying

with the Instrument Calibration definition also exists. Moreover,

the original TS bases submitted as part of the application for

Millstone Unit No. 1s Provisional Operating License (dated October

7, 1970) included recognition that instruments would be inoperable

during required functional test and calibrations. Thus, these

conditions were recognized in the original design bases and

constitute part of the licensing bases of the plant. NUREG-0123

provided specific time frames[,]...AOTs addressed in the table

notes[,] and specific action statements. Millstone Unit No. 1 AOT

values chosen are consistent with these values and less than those

approved in NUREG-1433 which had a more detailed study performed to

lengthen the AOT value.

The only action resulting from the proposed TS is to perform

testing as required by TS. Spurious signals during testing could

initiate equipment or plant transients but would be bounded by the

previous transient analysis. These tests do not subject the

instruments to any conditions beyond their design specifications and

are performed in accordance with approved testing standards. This

testing ensures equipment operability by identifying degraded

conditions, initiating corrective action and properly retesting

them. Therefore, the proposed TS do not involve a significant

reduction in a margin of safety.

The two-hour instrumentation AOT for the Air Ejector Off-Gas

System radiation monitors is slightly less restrictive than that

allowed by the NUREG-0123. Since this requirement was relocated from

NUREG-1433, there is no corresponding requirement for comparison.

These radiation monitors are arranged in a two-out-of-two logic;

therefore, both must trip to initiate the required action (closure

of the off-gas isolation valve to the main stack). This action,

however, is automatically delayed by 15 minutes. A high radiation

condition sensed by the monitor in service would provide sufficient

time to take corrective actions. Since a two-hour AOT is deemed

acceptable for instrumentation in system[s] such as the Reactor

Protection System and Emergency Core Cooling Systems, it is

appropriate to apply a two-hour AOT to these radiation monitors and

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

The addition of an allowable outage time of four hours for

Secondary Containment Integrity has negligible effect on accident

occurrence or consequences. Since the proposed change does not

involve the addition or modification of plant equipment, is

consistent with the intent of the existing Technical Specifications,

is consistent with the current industry practices as outlined in

NUREG-1433, (except for the deviation noted above), and is

consistent with the design basis of the plant and the accident

analysis, no action will occur that will involve a significant

reduction in a margin of safety.

Since the editorial items do not alter the meaning or intent of

any requirements, they do not affect 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: 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

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270. NRC Deputy Director: Phillip F. McKee

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

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

County, Minnesota

Date of amendment requests: July 28, 1995, as revised February 21,

1997

Description of amendment requests: The proposed amendments would

revise the Technical Specifications (TSs) to allow use of credit for

soluble boron in spent fuel pool criticality analyses. The licensee's

February 21, 1997, submittal is a revision to its original amendment

requests dated July 28, 1995. The generic methodology for crediting

soluble boron in spent fuel rack criticality analyses was approved

[[Page 14465]]

by the NRC on October 25, 1996. However, because of changes made to the

generic methodology as a result of comments from the NRC staff, it was

necessary for NSP to revise its original amendment requests. In

addition, the licensee has revised its request by eliminating the

proposed relocation of the spent fuel pool operating limits to the Unit

1 core operating limits report and will retain these limits in the TSs.

The licensee's original application for amendments was published in

the Federal Register on September 23, 1996, (61 FR 49800).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

increase in the probability or consequences of an accident

previously evaluated.

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

accident in the spent fuel pool when considering the presence of

soluble boron in the spent fuel pool water for criticality control.

The handling of the fuel assemblies in the spent fuel pool has

always been performed in borated water.

The criticality analysis showed the consequences of a fuel

assembly drop accident in the spent fuel pool are not affected when

considering the presence of soluble boron.

There is no increase in the probability of the accidental

misloading of spent fuel assemblies into the spent fuel pool racks

when considering the presence of soluble boron in the pool water for

criticality control. Fuel assembly placement will continue to be

controlled pursuant to approved fuel handling procedures and will be

in accordance with the Technical Specification spent fuel rack

storage configuration limitations. The addition of the spent fuel

pool storage configuration surveillance in proposed Specification

4.20 will provide increased assurance that a spent fuel pool

inventory verification will be completed in a timely manner after

completion of a fuel handling campaign in the spent fuel pool.

There is no increase in the consequences of the accidental

misloading of spent fuel assemblies into the spent fuel pool racks

because criticality analyses demonstrate that the pool will remain

subcritical following an accidental misloading if the pool contains

an adequate boron concentration. The proposed Technical

Specifications limitations will ensure that an adequate spent fuel

pool boron concentration will be maintained.

There is no increase in the probability of the loss of normal

cooling to the spent fuel pool water when considering the presence

of soluble boron in the pool water for subcriticality control since

a high concentration of soluble boron has always been maintained in

the spent fuel pool water.

A loss of normal cooling to the spent fuel pool water causes an

increase in the temperature of the water passing through the stored

fuel assemblies. This causes a decrease in water density which would

result in a decrease in reactivity when Boraflex neutron absorber

panels are present in the racks. However, since Boraflex is not

considered to be present, and the spent fuel pool water has a high

concentration of boron, a density decrease causes a positive

reactivity addition. However, the additional negative reactivity

provided by the proposed 1800 ppm boron concentration limit, above

that provided by the concentration required to maintain Keff

less than or equal to 0.95 (750 ppm), will compensate for the

increased reactivity which could result from a loss of spent fuel

pool cooling event. Because adequate soluble boron will be

maintained in the spent fuel pool water, the consequences of a loss

of normal cooling to the spent fuel pool will not be increased.

Therefore, based on the conclusions of the above analysis, the

proposed changes will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

a new or different kind of accident from any accident previously

analyzed.

Spent fuel handling accidents are not new or different types of

accidents, they have been analyzed in Section 14.5.1 of the Updated

Safety Analysis Report.

Criticality accidents in the spent fuel pool are not new or

different types of accidents, they have been analyzed in the Updated

Safety Analysis Report and in Criticality Analysis reports

associated with specific licensing amendments for fuel enrichments

up to 5.0 weight percent U-235.

The Prairie Island Technical Specifications currently contain

limitations on the spent fuel pool boron concentration. Current

Specification 3.8.E.2, which covers the storage of restricted fuel

assemblies in an unverified condition, and Specification 3.8.B.1.c

for the loading of fuel assemblies into a cask in the spent fuel

pool, contain requirements for spent fuel pool boron concentration.

The actual boron concentration in the spent fuel pool has always

been kept at a higher value for refueling purposes. New

Specification 3.8.E.2 establishes new boron concentration

requirements for the spent fuel pool water consistent with the

results of the new criticality analysis (Exhibit E [of the February

21, 1997, submittal]).

Since soluble boron has always been maintained in the spent fuel

pool water, and is currently required by Technical Specifications

under some circumstances, the implementation of this new requirement

will have little effect on normal pool operations and maintenance.

The implementation of the proposed new limitations on the spent fuel

pool boron concentration will only result in increased sampling to

verify boron concentration. This increased sampling will not create

the possibility of a new or different kind of accident.

Because soluble boron has always been present in the spent fuel

pool and is required by current Technical Specifications as

discussed above, a dilution of the spent fuel pool soluble boron has

always been a possibility. However, it was shown in the spent fuel

pool dilution evaluation (Exhibit D [of the February 21, 1997,

submittal]) that a dilution of the Prairie Island spent fuel pool

which could reduce the rack Keff to less than 0.95 is not a

credible event. Therefore, the implementation of new limitations on

the spent fuel pool boron concentration will not result in the

possibility of a new kind of accident.

Revised Specifications 3.8.E.1, 5.6.A.1.d and 5.6.A.1.e continue

to specify the requirements for the spent fuel rack storage

configurations, the only significant changes relate to the criteria

for determining the storage configuration. Since the proposed spent

fuel pool storage configuration limitations will be similar to those

currently in the Prairie Island Technical Specifications, the new

limitations will not have any significant effect on normal spent

fuel pool operations and maintenance and will not create any

possibility of a new or different kind of accident. Verifications

will continue to be performed to ensure that the spent fuel pool

loading configuration meets specified requirements.

As discussed above, the proposed changes will not create the

possibility of a new or different kind of accident. There is no

significant change in plant configuration, equipment design or

equipment. The accident analysis in the Updated Safety Analysis

Report remains bounding.

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

reduction in the margin of safety.

The Technical Specification changes proposed by this License

Amendment Request and the resulting spent fuel storage operating

limits will provide adequate safety margin to ensure that the stored

fuel assembly array will always remain subcritical. Those limits are

based on a plant specific criticality analysis (Exhibit E) performed

in accordance [with] the Westinghouse spent fuel rack criticality

analysis methodology described in Reference 4 [in Exhibit A of the

February 21, 1997, submittal].

While the criticality analysis utilized credit for soluble

boron, a storage configuration has been defined using a 95/95

Keff calculation to ensure that the spent fuel rack Keff

will be less than 1.0 with no soluble boron. Soluble boron credit is

used to offset uncertainties, tolerances and off-normal conditions

and to provide subcritical margin such that the spent fuel pool

Keff is maintained less than or equal to 0.95.

The loss of substantial amounts of soluble boron from the spent

fuel pool which could lead to exceeding a Keff of 0.95 has been

evaluated (Exhibit D) and shown to be not credible.

The evaluations in Exhibit D, which show that the dilution of

the spent fuel pool boron concentration from 1800 ppm to 750 ppm is

not credible, combined with the 95/95 calculation, which shows that

the spent fuel rack Keff will remain less than 1.0 when flooded

with unborated water, provide a level of safety comparable to the

conservative criticality analysis methodology required by References

1, 2 and 3 [in Exhibit A of the February 21, 1997, submittal].

[[Page 14466]]

Therefore, the proposed changes in this license amendment will

not result in a significant reduction in the plant's margin of

safety.

Based on the evaluation above, and pursuant to 10 CFR 50,

Section 50.91, Northern States Power Company has determined that

operation of the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by NRC

regulations in 10 CFR 50, Section 50.92.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

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

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

Obispo County, California

Date of amendment requests: February 14, 1997

Description of amendment requests: The proposed amendments would

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

Power Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance

frequencies from at least once every 18 months to at least once per

refueling interval (nominally 24 months) for 8 slave relay tests, 20

electrical system tests and 1 electrical TS Bases change, and 5

miscellaneous tests.

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 TS surveillance interval increase to 24 months do

not alter the intent or method by which the inspections, tests, or

verifications are conducted; do not alter the way any structure,

system, or component functions; and do not change the manner in

which the plant is operated.

The surveillance, maintenance, and operating histories indicate

that the equipment will continue to perform satisfactorily with

longer surveillance intervals. Few surveillance and maintenance

problems were identified. No problems have recurred following

identification of root causes and implementation of corrective

actions.

There are no known mechanisms that would significantly degrade

the performance of the evaluated equipment during normal plant

operation. All potential time related degradation mechanisms have

insignificant effects in the timeframe of interest (24 months +25

percent, or 30 months). Based on the past performance of the

equipment, the probability or consequences of accidents would not be

significantly affected by the proposed surveillance interval

increases.

Deletion of the phrase ``during shutdown'' for the applicable

electrical TS will not alter the intent or method by which the

inspections, tests, or verifications are conducted; nor alter the

way any structure, system, or component functions. DCPP has

administrative programs in place which require evaluation of risk

and suitability of surveillance and maintenance activities to ensure

that performance during plant operation does not adversely affect

safety.

The administrative change for one PORV TS regarding channel

calibration only maintains the existing surveillance frequency. This

revision does not alter the intent or method by which the

inspections, tests, or verifications are conducted; nor alter the

way any structure, system, or component functions.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

For the proposed TS changes involving surveillance interval

increases to 24 months, the surveillance and maintenance histories

indicate that the equipment will continue to effectively perform its

design function over the longer operating cycles. Additionally, the

increased surveillance intervals do not result in any physical

modifications, affect safety function performance or the manner in

which the plant is operated, or alter the intent or method by which

surveillance tests are performed. No problems have recurred

following identification of root causes and implementation of

corrective actions. All identified potential time related

degradations have insignificant effects in the timeframe of

interest. The proposed surveillance interval increases would not

affect the type of accident possible.

Deletion of the phrase during shutdown for the applicable

electrical TS does not result in any physical modifications, affect

safety function performance or the manner in which the plant is

operated, or alter the intent or method by which surveillance tests

are performed. DCPP has administrative programs in place which

require evaluation of risk and suitability of surveillance and

maintenance activities to ensure that performance during plant

operation does not adversely affect safety.

The administrative change for one PORV TS regarding channel

calibration only maintains the existing surveillance frequency. This

revision does not result in any physical modifications, affect

safety performance or the manner in which the plant is operated, or

alter the intent or method by which surveillance tests are

performed.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

For the proposed TS changes involving surveillance interval

increases to 24 months, evaluation of historical surveillance and

maintenance data indicates there have been few problems experienced

with the evaluated equipment. There are no indications that

potential problems would be cycle ength dependent or that potential

degradation would be significant for the timeframe of interest;

therefore, increasing the surveillance interval will have little, if

any, impact on safety. There is no safety analysis impact since

these changes will have no effect on any safety limit, protection

system setpoint, or limiting condition for operation, and there are

no hardware changes that would impact existing safety analysis

acceptance criteria. Safety margins would not be significantly

affected by the proposed surveillance interval increases.

Deletion of the phrase ``during shutdown'' for the applicable

electrical TS has no safety analysis impact since these changes will

have no effect on any safety limit, protection system setpoint, or

limiting condition for operation, and there are no hardware changes

that would impact existing safety analysis acceptance criteria. DCPP

has administrative programs in place which require evaluation of

risk and suitability of surveillance and maintenance activities to

ensure that performance during plant operation does not adversely

affect safety.

The administrative change for one PORV TS regarding channel

calibration only maintains the existing surveillance frequency. This

revision has no safety analysis impact.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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

Electric Company, P.O. Box 7442, San Francisco, California 94120

[[Page 14467]]

NRC Project Director: William H. Bateman

Portland General Electric Company, et al., Docket No. 50-344,

Trojan Nuclear Plant, Columbia County, Oregon

Date of amendment request: January 16, 1997, as supplemented

February 24, 1997.

Description of amendment request: The proposed amendment would

allow pre-operational testing and load handling of spent fuel transfer

and storage casks in the Trojan Fuel Building.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their 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 staff's review is

presented below:

The proposed changes would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

With the permanent cessation of operations at the Trojan Plant, the

number of potential accidents was reduced to those types of

accidents associated with the storage of irradiated fuel and

radioactive waste storage and handling. Additional events were

postulated for decommissioning activities due to the difference in

the types of activities that were to be performed. The postulated

accidents described in the Defueled Safety Analysis Report (DSAR)

are generally classified as: 1) radioactive release from a subsystem

or component, 2) fuel handling accident, and 3) loss of spent fuel

decay heat removal capability. The postulated events described in

the Decommissioning Plan are grouped as: 1) decontamination,

dismantlement, and materials handling events, 2) loss of support

systems (offsite power, cooling water, and compressed air), 3) fire

and explosions, and 4) external events (earthquake, external

flooding, tornadoes, extreme winds, volcanoes, lightning, toxic

chemical release). These types of accidents are discussed below.

Radioactive release from a subsystem or component involves failure

of a radioactive waste gas decay tank (WGDT) or failure of a chemical

and volume control system holdup tank (HUT). For a failure of a WGDT,

the radioactive contents are assumed to be principally noble gases

krypton and xenon, the particulate daughters of some of the krypton and

xenon isotopes, and trace quantities of halogens. For the failure of a

HUT, the assumptions were full power operation with 1-percent failed

fuel, 40 weeks elapsed since power operation, and 60,000 gallons of

120 deg.F liquid released over a 2-hour period. However, the WGDTs and

HUTs are no longer active and have been drained. Therefore, pre-

operational testing and load handling activities cannot increase the

probability of occurrence of a failure of a WGDT or HUT. Since the

failure of a WGDT or HUT is no longer credible, the consequences of

failure of a WGDT or HUT cannot significantly increase as a result of

pre-operational testing and load handling.

The fuel handling accident involves a stuck or dropped fuel

assembly that results in damage of the cladding of the fuel rods in one

assembly and the release of gaseous fission products. Pre-operational

testing and load handling do not involve the movement of irradiated

fuel. A dummy assembly will be used for fit-up testing. The fuel

handling equipment will be the same as previously analyzed with the

exception of special tools that may be used to manipulate the dummy

fuel assembly. These special tools will be similar in size and weight

to other tools used for underwater manipulation, and therefore, would

not present a new hazard. In addition, the same administrative controls

and physical limitations imposed on any fuel handling operation will be

used for pre-operational testing and load handling. Thus, there is no

increase in the probability of occurrence of a fuel handling accident

over what would be expected for any routine fuel handling operation. If

a dummy fuel assembly were dropped in the spent fuel pool, then only

one fuel assembly could be damaged. Therefore, the consequences of a

dummy fuel assembly drop would be the same as the consequences of the

analysis described in the DSAR. Therefore, the consequences of a dummy

fuel assembly drop are not significantly increased as a result of pre-

operational testing and load handling.

The loss of spent fuel decay heat removal capability involves the

loss of forced spent fuel cooling with and without concurrent spent

fuel pool (SFP) inventory loss. The only requirement to assume adequate

decay heat removal capability for the spent fuel is to maintain the

water level in the SFP so that the spent fuel assemblies remain covered

(i.e., the capability to makeup water to the SFP must be available when

required). The potential events that could result in a loss of spent

fuel decay heat removal capability include external events (explosions,

toxic chemicals, fires, ship collision with the intake structure, oil

or corrosive liquid spills in the river, cooling tower collapse,

seismic events, severe meteorological events), and internal events,

including SFP makeup water system malfunctions. Pre-operational testing

and load handling will not require the use of explosive materials,

toxic chemicals, or flammable materials. The probability of other

external events (e.g., cooling tower collapse) would be unaffected by

the pre-operational testing and load handling activities inside the

fuel building. Pre-operational testing and load handling activities

will not directly interface with the SFP makeup water systems, and

therefore could not affect their probability of failure. The safe load

path and handling height limitations will ensure that a load drop does

not adversely affect the SFP or makeup water systems. Therefore, there

is no significant increase in the probability of a loss of spent fuel

decay heat removal capability. There are no credible adverse

consequences of the loss of spent fuel decay heat removal as the DSAR

demonstrates that adequate time is available to establish a source of

makeup water to the SFP such that uncovering the fuel and an actual

loss of spent fuel cooling is not credible. The postulated events that

could affect the SFP (liner tear/breach and heavy load drop) do not

have a significant adverse effect. In addition, establishment of the

makeup water path and recovery of spent fuel cooling would not be

affected because postulated off-normal events and accidents could not

affect the capability to provide makeup water to the SFP by various

water sources. Therefore, pre-operational testing and load handling

cannot significantly increase the consequences of the loss of spent

fuel decay heat removal.

The events postulated in the Decommissioning Plan are similar to

the DSAR with the exception of decontamination, dismantlement, and

materials handling events. Decontamination events involve gross liquid

leakage from in-situ decontamination equipment or accidental spraying

of liquids containing concentrated contamination. Dismantlement events

include segmentation of components and structures, or removal of

concrete by rock splitting, explosives, or electric and/or pneumatic

hammers. Dismantlement events potentially result in airborne

contamination. Materials handling events involve dropping contaminated

components, concrete rubble, or filters or packages of particulate

materials. Pre-operational testing and load handling activities are

material handling activities and are therefore, within the bounds of

the existing analysis. Therefore, the probability and consequences of

decontamination, dismantlement, and materials handling events would not

be significantly increased.

[[Page 14468]]

Based on the above, the pre-operational testing and load handling

activities do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated. As

described in the licensee's safety evaluation of the proposed pre-

operational testing and load handling activities, no types of off-

normal events/accidents were determined to have radiological

consequences greater than currently evaluated in the DSAR and

Decommissioning Plan.

The postulated dummy fuel assembly drop is considered the same type

or kind of event as the previously analyzed fuel handling accident,

mainly because the initiator for this postulated event is the same

(i.e., a (non-specified) failure of the fuel handling equipment or the

fuel handling bridge crane. During pre-operational testing and load

handling, a dummy fuel assembly could be dropped in the SFP or the cask

loading pit. As the cask loading pit is similar in construction to the

SFP and the cask loading pit will be flooded with borated water of the

same concentration as the SFP, the differences between the two events

are negligible and the two events may be considered the same type or

kind of accident. Therefore the dummy fuel assembly drop is not a new

or different type or kind of accident.

The postulated transfer cask drop or mishandling event is similar

to a materials handling event. Therefore, the consequences of a

transfer cask drop or mishandling event would not represent a new or

different type or kind of accident.

Based on the above, the pre-operational testing and load handling

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

accident.

The proposed changes do not involve a significant reduction in the

margin of safety. The Trojan Permanently Defueled Technical

Specifications (PDTS) contain four limiting conditions of operation

that address SFP water level, SFP boron concentration, SFP temperature,

and SFP load restrictions. These PDTS will remain in effect as long as

spent fuel is stored in the SFP, which is in accordance with their

applicability statements. The pre-operational testing and load handling

activities will not affect these PDTS or their bases.

The cask loading pit (CLP) is immediately adjacent to the SFP. The

gate between the CLP and the SFP may be opened to allow a dummy fuel

assembly to moved from the spent fuel storage racks in the SFP to the

basket in the CLP. Opening the gate will allow free exchange of water

between the CLP and the SFP. The water in the CLP must be at

essentially the same level, boron concentration, and temperature as the

SFP prior to the first opening of the gate to ensure that the limited

conditions of operation are continuously satisfied for the SFP.

Therefore, the CLP will be initially filled to about the same level as

the SFP with water that is about the same boron concentration and

temperature as the SFP. With these precautions, the limiting conditions

of operation for SFP level, boron concentration, and temperature will

be continuously maintained and the margin of safety will be unaffected.

Pre-operational testing and load handling activities will involve

lifting and moving heavy loads (e.g., transfer casks). Loads that will

be carried over fuel in the SFP racks and the heights at which they may

be carried will be limited in accordance with LCO 3.1.4, ``Spent Fuel

Pool Load Restrictions,'' in such a way as to preclude impact energies

over 240,000 in-lbs. With this precaution, the limiting condition of

operation pertaining to load restrictions over the SFP will be

satisfied for fuel stored in the SFP racks and the margin of safety

will be unaffected. The safe load path for heavy loads being lifted and

moved outside the SFP will be located sufficiently far from the SFP as

to not have an adverse effect on the SFP in the unlikely event of a

load drop. In addition, the mechanical stops and electrical interlocks

on the fuel building overhead crane will provide additional assurance

that heavy loads are not carried over the fuel in the SFP racks.

Based on the above, the pre-operational testing and load handling

activities will not reduce the margin of safety.

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: Branford Price Millar

Library, Portland State University, 934 S.W. Harrison Street, P.O. Box

1151, Portland, Oregon 97207

Basis for proposed no significant hazards consideration

determination:

Attorney for licensees: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204

NRR Project Director: Seymour H. Weiss

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

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

Date of amendment request: January 2, 1997

Description of amendment request: The proposed amendment would

allow a change to the current functional testing frequency for

Inservice Inspection of American Society of Mechanical Engineers Code

Class 1, 2, and 3 pumps and valves from the current monthly to a

quarterly testing frequency.

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 proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: Operation of Indian Point 3 in accordance with the

proposed license does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes involve no hardware changes, no changes to

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

existing structures. 10 CFR 50.55a(g) requires that safety related

components (e.g. - pumps and valves) be tested according to the

requirements of Section XI of the American Society of Mechanical

Engineers (ASME) Boiler and Pressure Vessel Code (Code) and

applicable addenda. The revision of functional test frequencies for

pumps and valves, which are categorized as Code Class 1, 2, or 3,

from a monthly to a quarterly test interval is consistent with NRC

guidance provided in NUREG-1366 and in accordance with recommended

test intervals in the ASME Code. These changes will reduce component

degradation resulting from unnecessary tests and provide better

system availability from not having to remove a system/component

from operability while performing a surveillance. Such changes will

not alter the probability or consequences of any previously analyzed

accidents.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response: The proposed change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

The proposed changes are procedural in nature concerning the

functional testing frequencies of pumps and valves that have

historically shown a high percentage of successfully meeting

surveillance requirements. The methodology of testing these pumps

and valves will remain unchanged. The proposed changes, while

slightly increasing the possibility of an

[[Page 14469]]

undetected pump or valve defect, will not create a new or

unevaluated accident or operating condition.

(3) Does the proposed license amendment involve a significant

reduction in a margin of safety?

Response: The proposed license amendment does not involve a

significant reduction in a margin of safety.

The proposed changes are in accordance with recommendations

provided by the NRC regarding the improvement of Technical

Specifications. These changes will result in the perpetuation of

current safety margins while reducing the testing burden and

decreasing equipment degradation.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

New York, New York 10019.

NRC Project Director: S. Singh Bajwa, Acting Director

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: February 11, 1997

Description of amendment request: The proposed change to Hope Creek

Technical Specification (TS) Sections 3/4.8.1 ``A.C. Sources,'' 6.8

``Procedures and Programs,'' and the Bases for Section 3/4.8,

``Electrical Power Systems,'' would include: 1) the relocation of

existing surveillance requirements related to diesel fuel oil

chemistry; 2) the introduction of a new program under TS 6.8.4.e,

``Diesel Fuel Oil Testing Program;'' 3) revisions to the TS Bases for

Section 3/4.8 to incorporate information associated with the TS

changes; and 4) editorial changes to implement required corrections.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes involve: 1) no hardware changes; 2) no

significant changes to the operation of any systems or components in

normal or accident operating conditions; and 3) no changes to

existing structures, systems or components. Therefore these changes

will not increase the probability of an accident previously

evaluated.

Establishment of [Emergency Diesel Generator] EDG fuel oil

testing requirements in TS 6.8.4.e is a change that is consistent

with changes made in the improved STS [Standard Technical

Specifications] as contained in Specification 5.5.10 of that

document. These changes establish a new requirement to test for

particulates in the EDG fuel oil, but establish a 92 day test

frequency (as opposed to 31 days in the improved STS) and a 3.0

micron acceptance criteria (as opposed to 0.8 micron in the improved

STS) for particulate testing. [Public Service Electric and Gas

Company] PSE&G concludes that these changes are acceptable based

upon past EDG fuel oil tests for particulates and acceptable

performance of the EDG with 5.0 micron filters. In addition, PSE&G

will utilize more objective test criteria for water and sediment in

the EDG fuel oil than established by the ``clear and bright''

acceptance criteria contained in the improved STS.

Since the EDG fuel oil will still: 1) meet all of the

requirements established for fuel oil specified in the improved STS;

and 2) retain the capability to mitigate the consequences of

accidents described in the [Hope Creek Generating Station] HC Safety

Analysis Report, the proposed changes were determined to be

justified. Based on established fuel oil quality history, the

proposed testing methods and frequencies will not significantly

decrease confidence in fuel oil quality and EDG operability, nor

will they have any negative effect on established plant practices in

regards to the testing of EDG fuel oil. Therefore, these changes

will not involve a significant increase in the consequences of an

accident previously evaluated.

The revisions proposed to the TS Bases are being made to provide

additional information supporting the proposed EDG TS. With the

approval of the proposed TS changes, the associated Bases changes

would be editorial in nature. Therefore, these changes will not

involve a significant increase in the consequences of an accident

previously evaluated.

In addition, the proposed change to [Limiting Condition for

Operation] LCO 3.8.1.1, ACTION c., is considered to be editorial in

nature and will not result in a significant increase in the

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The HC EDGs are designed to mitigate the consequences of

accidents by providing electrical power to safety-related equipment.

Failure of the EDGs are not considered to initiate any of the

accidents described in the HC Safety Analysis Report. The proposed

changes concern fuel oil system surveillances and testing frequency.

The proposed changes will not adversely impact the operation of any

safety related component or equipment. Since the proposed changes

involve: 1) no hardware changes; 2) no significant changes to the

operation of any systems or components; and 3) no changes to

existing structures, systems or components, there can be no impact

on the occurrence of any accident. Furthermore, there is no change

in plant testing proposed in this change request which could

initiate an event. Therefore, these changes will not create the

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

previously evaluated.

In addition, the proposed change to LCO 3.8.1.1, ACTION c., is

considered to be editorial in nature and will not result in a new or

different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Establishment of EDG fuel oil testing requirements in TS 6.8.4.e

is a change that is consistent with changes made in the improved

STS. The proposed changes address: 1) how EDG fuel oil quality is to

be determined; 2) how frequently this determination is to be

performed; and 3) how to control the process for determining fuel

oil acceptability and resultant EDG operability. With the exception

of particulate testing (which is being added) all acceptance

criteria for fuel oil testing remain unchanged. Based on historical

data, EDG fuel oil quality will not be adversely affected or

impacted by the proposed changes. Therefore, the proposed amendment

does not involve any significant reduction in a safety margin.

The revisions proposed to the TS Bases are being made to provide

additional information supporting the proposed EDG TS. With the

approval of the proposed TS changes, the associated Bases changes

would be editorial in nature. Therefore, these changes will not

involve a significant reduction in a safety margin.

In addition, the proposed change to LCO 3.8.1.1, ACTION c., is

considered to be editorial in nature and will not involve a

significant reduction in a safety margin.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear

Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: October 23, 1996, January 31, February

10 and 24 and March 11, 1997.

Description of amendment request: The proposed amendment would

revise

[[Page 14470]]

the Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications to

increase the enrichment and storage capacity of the spent fuel pool

racks. The proposed modification increases the (Watts Bar Nuclear

Plant) WBN spent fuel storage capacity from 484 fuel assemblies to 1835

fuel assemblies. The initial enrichment of the fuel to be stored in the

spent fuel storage racks will be increased from 3.5 weight percent

(wt%) to 5.0 wt%. This modification would also change the spacing of

stored fuel assembly center-to-center spacing from a nominal 10.72

inches to 10.375 inches in 24 PaR flux trap rack modules and 8.972

inches in ten smaller burnup credit rack modules to be installed

peripherally along the south and west pool walls and in a single 15 x

15 burnup credit rack to be installed in the cask pit.

In addition to the above proposed revisions, two limiting

conditions for operation will be added to require that the combination

of initial enrichment and burnup of each spent fuel assembly to be

stored is in the acceptable region and to require boron concentration

of the cask pit to be greater than or equal to 2000 parts per million

(ppm) during fuel movement in the flooded cask pit. As an added

protection to the fuel stored in the cask pit area, the Technical

Requirements Manual (TRM) is being revised to require that an impact

shield be in place over the fuel when heavy loads are moved near or

across the cask pit area.

The WBN Unit 1 Technical Specification Bases and the TRM would be

revised to support these changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The Nuclear Regulatory Commission has provided standards for

determining whether a significant hazards consideration exists (10

CFR 50.92(c)). A proposed amendment to an operating license for a

facility involves no significant hazards consideration if 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. Each standard is discussed below for the proposed

amendment.

(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 following potential scenarios were considered:

1. A spent fuel assembly drop.

2. Drop of the transfer canal gate or the cask pit divider gate.

3. A seismic event.

4. Loss-of-cooling flow in the spent fuel pool.

5. Installation activities.

The effect of additional spent fuel pool storage cells fully

loaded with fuel on the first four potential accident scenarios

listed above has been considered. It was concluded that after

installation activities have been completed, the presence of

additional fuel in the pool does not increase the probability of

occurrence of these four events. Also, based on evaluations of bulk

pool temperature, rack seismic responses, and refueling accidents,

it is reasonable to conclude that there is no significant increase

in the consequences of these events after installation is complete

(See Reference 1). During the installation activities, the following

considerations support a conclusion that neither the probability or

consequences of these four scenarios would be significantly

increased.

A spent fuel assembly cannot be dropped during installation of

the 24 Programmed and Remote System Corporation (PaR) flux trap rack

modules because this activity will take place before the end of

operating cycle one and there will be no spent fuel in the WBN pool

to be moved or shuffled. Before installing the ten smaller burnup

credit racks in the pool, some fuel will be moved to create a three

foot lateral free zone clearance from stored fuel. This would

involve a one-time movement of an estimated maximum of 225 fuel

assemblies, which is less that half the fuel movements during one

refueling outage. This does not significantly increase the

probability of dropping a fuel assembly, particularly when the many

administrative controls and physical limitations imposed on fuel

handling operations are considered. The fuel handling system

consists of equipment and structures utilized for safely

implementing refueling operations in accordance with requirements of

General Design Criteria 61 and 62 of 10 CFR 50, Appendix A. The

radiological dose consequences of dropping a 5.0 wt% fuel assembly

are different from the previous FSAR [Final Safety Analysis Report]

evaluation for the 3.5 wt% fuel assembly. The Beta and Gamma doses

decrease and the maximum thyroid dose increase is less than 9%.

Therefore, the change in calculated dose values is insignificant and

remains well within regulatory guidelines.

It may be necessary to move the transfer canal gate and the cask

pit divider gate between their gated and stored positions during

installation of the burnup credit ``baby'' rack modules along the

south and west walls. During rack installation, the previously

mentioned three foot lateral free zone clearance to stored fuel

would exist. Therefore, no heavy load would be carried directly over

irradiated fuel during installation of the racks. There are numerous

design features which comply with NUREG-0612 to preclude these gates

from dropping on spent fuel. These features include design of the

lifting devices, design of the crane, and use of written procedures.

Also, the evaluation results for a gate drop on the racks indicates

that permanent damage to a fuel storage cell is limited to a maximum

depth of less than six inches below the top of the rack with no

effect on the subcriticality of fuel stored in adjacent cells. Based

on the foregoing, it is reasonable to conclude that gate handling

during the installation of the ``baby'' racks would not involve a

significant increase in the probability or consequences of an

accident.

The probability of a seismic event is not related to

installation activities. The worst consequence resulting from a

seismic event during installation activities would occur during

handling of a rack. The consequences would be insignificant because

the Auxiliary Building crane is seismically qualified and both

handling equipment and operations meet the criteria of NUREG-0612.

Nevertheless, if the seismic event resulted in a rack drop, the

consequences are insignificant, i.e., localized damage to the pool

liner and a minor leak rate which would be small in comparison to

available installed makeup capacity. The cooling and shielding of

the spent fuel would remain unaffected. Also the racks being moved

are empty during installation and therefore, the criticality

consequences of seismic events are bounded by evaluations for loaded

racks.

Rack installation activities cannot cause an accidental loss-of-

cooling flow in the spent fuel pool. The vital components of the

spent fuel pool cooling and cleanup system (SFPCCS) are not located

proximate to the pool installation activities. Coolant flow may be

deliberately curtailed to facilitate installation of the ``baby''

racks directly beneath the discharge piping in the southwest corner

of the pool. The effects of such an action would be readily

minimized and made inconsequential during the detailed installation

planning phase by selecting a time when decay heat input from stored

fuel is relatively constant. Also careful preplanning of the work

would minimize out-of-service time and provide for intermittent

coolant flow restart, if necessary, to maintain acceptable bulk

coolant temperatures. Similarly, the effect of an independently

initiated loss-of-coolant flow incident on reracking activities can

be easily accommodated by stopping work, as necessary, to mitigate

any adverse effects on the installation process. The consequences of

loss-of-cooling flow in the spent fuel pool during installation are

bounded by the analysis in Chapter 5 of the report which includes

the situation in which ``baby'' racks and the 15 x 15 cask pit rack

are installed, and the pool is filled to capacity with spent fuel.

With regard to the actual installation activities, the existing

WBN TRM prohibits loads in excess of 2059 pounds from travel over

fuel assemblies in the storage pool and requires the associated

crane interlocks and physical stops be periodically demonstrated

operable. During installation, racks and associated handling tools

will be moved over the spent fuel pool, however there will be no

fuel in the pool when the 24 flux trap rack modules are installed. A

three foot lateral free zone clearance from stored spent fuel

[[Page 14471]]

will be maintained during installation of the ten smaller burnup

credit rack modules. Installation work in the spent fuel pit area

will be controlled and performed in strict accordance with specific

written instructions.

NUREG-0612 states that in lieu of providing a single failure-

proof crane system, the control-of-heavy-loads guidelines can be

satisfied by establishing that the potential for a heavy load drop

is extremely small. Storage rack movements to be accomplished with

the WBN Auxiliary Building crane will conform with NUREG-0612

guidelines in that the probability of a drop of a storage rack is

extremely small. The crane has a tested capacity of 125 tons. The

maximum weight of any existing, replacement, or new storage rack and

its associated handling tool is less than 20 tons. Therefore, there

is ample safety factor margin for movements of the storage racks by

the Auxiliary Building crane. Special lifting devices, which have

redundancy or a rated capacity sufficient to maintain adequate

safety factors, will also be utilized in the movements of the

storage racks. In accordance with NUREG-0612, Appendix B, the safety

margin ensures that the probability of a load drop is extremely low.

Future load travel over fuel stored in a rack specifically

designed for the cask loading area of the cask pit will be

prohibited unless an impact shield, which has been specifically

designed for this purpose, is covering the area. Loads that are

permitted when the shield is in place must meet analytically

determined weight, travel height, and cross-sectional area criteria

that preclude penetration of the shield. A Technical Requirement

(TR) has been proposed that incorporates the previously mentioned

load criteria.

Also a rack change-out sequence is being developed that

addresses removal of the existing racks, movement of the new racks

into the Auxiliary Building, initial staging on the refueling floor,

and final installation in the pool. The change-out sequence

objectives include establishing lift heights, travel distances, and

number of lifts to be as low as reasonably achievable. Accordingly,

it is concluded that the proposed installation activities will not

significantly increase the probability of a load-handling accident.

The consequences of a load-handling accident are unaffected by the

proposed installation activities.

The consequences of a spent fuel assembly drop were evaluated,

and it was determined that the racks will not be distorted such that

the racks would not perform their safety function. The criticality

acceptance criterion, Keff less than or equal to 0.95, is not

violated, and the calculated doses are well within 10 CFR Part 100

guidelines. The radiological consequences of the fuel assembly drop

accident evaluated for WBN, have changed, however, the changes do

not involve a significant increase in consequences and are well

within the 10 CFR 100 requirements.

A TRM change has been proposed that would permit the transfer-

canal gate and the divider gate for the cask pit to travel over fuel

assemblies in the spent fuel pool during movement between their

gated and stored position. Rack damage is restricted to an area

above the active fuel region, therefore, neither criticality nor

radiological concerns exist.

The consequences of a seismic event have been evaluated. The

replacement racks are designed and fabricated and the new racks will

be fabricated to meet the requirements of applicable portions of the

NRC regulatory guides and published standards. Design margins have

been provided for rack tilting, deflection, and movement such that

the racks do not impact each other or the spent fuel pool walls in

the active fuel region during the postulated seismic events. The

free-standing racks will maintain their integrity during and after a

seismic event. The fuel assemblies also remain intact and therefore

no criticality concerns exist.

The spent fuel pool system is a passive system with the

exception of the fuel pool cooling train and heating, ventilating,

and air-conditioning (HVAC) equipment. Redundancies in the cooling

train and HVAC hardware are not reduced by the planned fuel storage

modification. The potential increased heat load resulting from any

additional storage of spent fuel is well within the existing system

cooling capacity. Therefore, the probability of occurrence or

malfunction of safety equipment leading to the loss-of-cooling flow

in the spent fuel pool is not significantly affected. Furthermore,

the consequences of this type incident are not significantly

increased from previously evaluated cooling system loss of flow

malfunctions. Thermal-hydraulic scenarios assume the reracked pool

is approximately 90% full with spent fuel assemblies. From this

starting point, the remaining storage capacity is utilized by

analyzing both normal and unplanned full core off loads using

conservative assumptions and previously established methods.

Calculated values include maximum pool water bulk temperature,

coincident maximum pool water local temperature, the maximum fuel

cladding temperature, time-to-boil after loss-of-cooling paths, and

the effect of flow blockage in a storage cell.

Although the proposed modification increases the pool heat load,

results from the above analyses yield a maximum bulk temperature

less than 160 degrees Fahrenheit which is below the bulk boiling

temperature. Also the maximum local water temperature is below

nucleate boiling condition values. Associated results from

corresponding loss-of-cooling evaluations give minimums of 5.3 hours

before boiling begins and 45 hours before the pool water level drops

to the minimum required for shielding spent fuel. This is sufficient

time to begin utilization of available alternate sources of makeup

cooling water. Also, the effect of the increased thermal loading on

the pool structure, associated cooling system, and components was

evaluated and determined to establish an acceptable design basis

with the new storage configuration. No modifications were necessary

because of the increased temperature.

(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 analyzed.

The proposed modification has been evaluated in accordance with

the guidance of the NRC position paper entitled, ``OT Position for

Review and Acceptance of Spent-Fuel Storage and Handling

Applications'', appropriate NRC regulatory guidelines; appropriate

NRC standard review plans; and appropriate industry codes and

standards. Proven analytical technology was used in designing the

planned fuel storage expansion and will be utilized in the

installation process. Basic reracking technology has been developed

and demonstrated in applications for fuel pool capacity increases

that have already received NRC staff approval.

Proposed TSs for the spent fuel storage racks use burnup credit

and fuel assembly administrative placement restrictions for

criticality control. These restrictions are described in the

proposed change to the design features section of the TSs by

reference to the Spent Fuel Pool Modifications report. Additional

evaluations were required to ensure that the criticality criterion,

keff less than 0.95, is maintained. These include evaluation

for the abnormal placement of unirradiated (fresh) fuel assemblies

of 5.0 wt% enrichment into a storage cell location designed for

lower enrichment or irradiated fuel. Soluble boron, for which credit

is permitted under these abnormal conditions, ensures that

reactivity is maintained substantially less than the design

requirement. For example, if the PaR flux trap racks are

inadvertently all loaded with fresh assemblies of the maximum 5.0

wt% fuel instead of observing the 3.8 wt% and 6.75 MWD/KgU controls,

the worth of the 2000 ppm borated water is sufficient to lower the

keff of the storage racks to 0.83. The existing and proposed

TSs require boron concentration in the pool and cask pit to be more

than or equal to 2000 ppm during fuel movement. An analytical

determination of the reactivity worth of 2000 ppm borated water in

the spent fuel storage pool predicted the change in keff to be

approximately 17 percent keff. Although no credit for soluble

boron was proposed in the TSs, it was also determined by an

independent calculation that a minimum concentration of 520 ppm

soluble boron allows the unrestricted storage of 5.0 wt% enriched

fuel in the PaR flux trap racks.

The Holtec-designed peripheral ``baby'' racks and the 15 x 15

racks in the cask loading area can safely and conservatively store

fuel of 5 wt% initial enrichment burned to 41 MWD/kgU or lower

enriched fuel with lower burnup, i.e., fuel of equivalent

reactivity. Evaluations have confirmed that, for the abnormal

placement of a fresh fuel assembly of 5.0 wt% in these racks, the

criticality criterion is maintained with the existing and proposed

TS requirements of 2000 ppm soluble boron.

Although these changes required addressing additional aspects of

a previously analyzed accident, the possibility of a previously

unanalyzed accident is not created.

The impact shield design together with its attendant

administrative controls and NUREG-0612 heavy load lift compliance,

renders the possibility of a heavy load drop

[[Page 14472]]

on fuel as not credible in accordance with the NUREG-0612 single-

failure-proof criteria. Accordingly, since this particular part of

the proposed reracking modification is not a change that could

malfunction by a new single failure, the movement of heavy loads

over the cask pit does not create the possibility of a new or

different kind of accident.

It is therefore concluded that the proposed reracking does not

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

any previously analyzed.

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

amendment would not involve a significant reduction in a margin of

safety.

The design and technical review process applied to the reracking

modification included addressing the following areas:

1.

Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material, and structural considerations.

The established acceptance criterion for criticality is that the

neutron multiplication factor shall be less than or equal to 0.95,

including all uncertainties. The results of the criticality analyses

for the rack designs demonstrate that this criterion is satisfied.

The methods used in the criticality analysis conform to the

applicable portions of NRC guidance and industry codes, standards,

and specifications. In meeting the acceptance criteria for

criticality in the spent fuel pool and the cask loading area, such

that keff is always less than 0.95 at a 95/95 percent

probability tolerance level, the proposed amendment does not involve

a significant reduction in the margin of safety for nuclear

criticality.

Conservative methods and assumptions were used to calculate the

maximum fuel temperature and the increase in temperature of the

water in the spent fuel pit area. The thermal-hydraulic evaluation

used methods previously employed. The proposed storage modification

will increase the heat load in the spent fuel pool, but the

evaluation shows that the existing spent fuel cooling system will

maintain the bulk pool water temperature at or below 160 degrees

Fahrenheit. Thus it is demonstrated that the worst-case peak value

of the pool bulk temperature is considerably lower than the bulk

boiling temperature. Evaluation also shows that maximum local water

temperatures along the hottest fuel assembly are below the nucleate

boiling condition value. Thus, there is no significant reduction in

the margin of safety for thermal hydraulic or spent fuel cooling

considerations.

The mechanical, material, and structural design of the spent

fuel racks is in accordance with applicable portions of NRCs

position in ``OT Position for Review and Acceptance of Spent-Fuel

Storage and Handling applications,'' dated April 14, 1978 (as

modified January 18, 1979), as well as other applicable NRC guidance

and industry codes. The primary safety function of the spent fuel

racks is to maintain the fuel assemblies in a safe configuration

through normal and abnormal loading conditions. Abnormal loadings

that have been evaluated with acceptable results and discussed

previously include the effect of an earthquake and the impact

because of the drop of a fuel assembly. The rack materials used are

compatible with the fuel assemblies and the environment in the spent

fuel pool. The structural design for the new racks provides tilting,

deflection, and movement margins such that the racks do not impact

each other or the spent fuel pit walls in the active fuel region

during the postulated seismic events. Also the spent fuel assemblies

themselves remain intact and no criticality concerns exist. In

addition, finite element analysis methods were used to evaluate the

continued structural acceptability of the spent fuel pit. The

analysis was performed in accordance with ``Building Code

Requirements for Reinforced Concrete,'' (ACI 318-63,77). Therefore,

with respect to mechanical, material, and structural considerations,

there is no significant reduction in a margin of safety.

Summary

Based on the above analysis, TVA has determined that operation

of WBN, in accordance with the proposed amendment, would not: (1)

involve a significant increase in the probability of consequences of

an accident previously evaluated, (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. Therefore, operations of WBN in accordance with the proposed

amendments as described do not involve significant hazard

considerations as defined in 10 CFR 50.92 and that the criteria of

10 CFR 50.91 have accordingly been met.

TVA has also reviewed the NRC examples of licensing amendments

considered not likely to involve significant hazards considerations

as provided in the final adoption of 10 CFR 50.92 published on page

7751 of the Federal Register, Volume 51, No. 44, March 6, 1986.

Example (X) provides four criteria that, if satisfied by a reracking

request, indicate that it is likely no significant hazards

considerations are involved. The criteria and how TVAs amendment

request for WBN complies are indicated below.

Criterion (1):

The storage expansion method consists of either replacing

existing racks with a design that allows closer spacing between

stored spent fuel assemblies or replacing additional racks of the

original design on the pool floor if space permits.

Proposed Amendment:

The WBN reracking involves replacing the existing racks with a

design that allows slightly closer spacing between stored fuel

assemblies and also provides additional rack storage on the pool

floor where space permits.

Criterion (2):

The storage expansion method does not involve rod consolidation

or double tiering.

Proposed Amendment:

The WBN racks are not double tiered, and the racks will sit on

the floor of the spent fuel pool. Additionally, the amendment

application does not involve consolidation of spent fuel.

Criterion (3):

The keff of the pool is maintained less than or equal to

0.95.

Proposed Amendment

The design of the spent fuel racks contains a neutron absorber,

Boral, to allow close storage of spent fuel assemblies while

ensuring that the keff remains less than 0.95 under normal

operating conditions with unborated water in the pool and less than

0.95 under abnormal conditions with soluble boron in the pool.

Criterion (4):

No new technology or unproven technology is utilized in either

the construction process or the analytical techniques necessary to

justify the expansion.

Proposed Amendment:

The construction processes and analytical techniques used in the

fabrication and design are substantially the same as those of

numerous other rack installations, Thus, no new or unproven

technology is utilized in the construction or analysis of the high

density, spent fuel racks at WBN. TVA's contractor, Holtec

International, has previously supplied licensable racks of several

similar design for about 10 other reracking projects

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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

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

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: November 26, 1996

Description of amendment request: The proposed changes would

eliminate the records retention requirements from the administrative

section of the Technical Specifications (TS) in accordance with NRC

Administrative Letter 95-06, ``Relocation of Technical Specifications

Administrative Controls Related to Quality Assurance.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Specifically, operation of the Surry... Power [Station] in

accordance with the

[[Page 14473]]

proposed Technical Specifications changes will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated. The proposed

administrative changes do not affect equipment or its operation.

Therefore, the likelihood that an accident will occur is neither

increase nor decreased by relocating record retention requirements

from the Technical Specifications to the Operational Quality

Assurance Program. This TS change will not impact the function or

method of operation of plant equipment. Thus, a significant increase

in the probability of a previously analyzed accident does not result

due to this change. No systems, equipment, or components are

affected by the proposed changes. Thus, the consequences of any

accident previously evaluated in the UFSAR [Updated Final Safety

Analysis Report] are not increased by this change.

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

accident from any accident previously evaluated. The proposed change

does not alter the design or operations of the physical plant. Since

record retention requirements are administrative in nature, a change

to these requirements does not contribute to accident initiation, an

administrative change related to this activity does not produce a

new accident scenario or produce a new type of equipment

malfunction. [These] changes do not alter any existing accident

scenarios. The proposed administrative change does not affect

equipment or its operation, and, thus, does not create the

possibility of a new or different kind of accident. Therefore, the

proposed change does not create the possibility of a new or

different kind of accident.

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

Section 6.0 of the...Surry Technical Specifications does not have a

basis description. The proposed administrative change does not

affect equipment or its operation, and, thus, does not involve any

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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

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

Virginia 23219.

NRC Project Director: Mark Reinhart, Acting

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert

Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application for amendment: July 31, 1997, as supplemented

February 13, 1997.

Brief description of amendment: The proposed amendment would revise

the Technical Specifications to reduce the minimum Reactor Coolant

System total flow rate from 370,000 gpm to 340,000 gpm. The proposed

changes are necessary to support a larger number of plugged steam

generator tubes for future operating cycles.

Date of publication of individual notice in Federal Register:

February 26, 1997 (62 FR 8780)

Expiration date of individual notice: March 28, 1997

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: February 14, 1997

Brief description of amendment: The proposed amendment would revise

the Technical Specifications to permit a one-time extension of the

current steam generator tube inservice inspection cycle. Date of

publication of individual notice in Federal Register: March 4, 1997 (62

FR 9816)

Expiration date of individual notice: March 28, 1997

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: February 17, 1997

Brief description of amendment: Changes to Technical Specification

to implement 10 CFR 50, Appendix J Option B relating to containment

leakage tests.

Date of publication of individual notice in the Federal Register:

February 28, 1997 (62 FR 9214).

Expiration date of individual notice: March 31, 1997

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: February 14, 1997

Brief description of amendment request: The proposed amendment

would revise Technical Specification (TS) Section 3/4.5.2, ``Emergency

Core Cooling Systems, ECCS Subsystems - Tavg more than or equal to

280 deg.F.'' Surveillance requirement 4.5.2.f would be modified to

state that opening and closing of the inspection port on the watertight

enclosure for the decay heat valve pit would not require this

surveillance procedure to be performed. The applicable TS bases would

also be changed. Date of publication of individual notice in Federal

Register: February 26, 1997 (62 FR 8783) Expiration date of individual

notice: March 28, 1997

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

[[Page 14474]]

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.

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

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

County, Maryland

Date of application for amendments: August 1, 1996

Brief description of amendments: The amendments modify the

Technical Specifications requirements to allow use of blind flanges

during Modes 1-4 in the Calvert Cliffs 1 and 2 Containment Purge system

instead of the two outboard 48-inch isolation valves. Date of issuance:

March 7, 1997

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

end of the 1998 refueling outage for Unit 1; by the end of the 1997

refueling outage for Unit 2.

Amendment Nos.: 221 and 197

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

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

FR 47975) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated March 7, 1997 No significant

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

North Carolina

Date of application for amendment: December 30, 1996

Brief description of amendment: The amendment revises chemistry

data for TS Figures 3.4-2 and 3.4-3 and the associated Bases.

Date of issuance: March 7, 1997

Effective date: March 7, 1997

Amendment No.: 68

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

Technical Specifications

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4342) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 7, 1997. No significant hazards

consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: September 20, 1996, as

supplemented January 21, 1997.

Brief description of amendments: The amendments would update the

pressure- temperature cures contained in the Dresden and Quad Cities

Technical Specifications to 22 Effective Full Power Years. Date of

issuance: February 28, 1997 Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 153, 148, 172 and 168

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66703). The January 21, 1997, submittal provided additional

clarifying information that did not change the original proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated February 28, 1997 No significant hazards consideration

comments received: No

Local Public Document Room location: for Dresden, Morris Area

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

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

Illinois 61021.

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: December 6, 1996

Brief description of amendments: The amendments would change the

Technical Specification (TS) by allowing a single control rod to be

moved when the plant is in the Hot Shutdown or Cold Shutdown condition

provided that the one-rod-out interlock is Operable and the reactor

mode switch is in the refuel position.

Date of issuance: March 4, 1997

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

Amendment Nos.: 154, 149, 173, 169

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 15, 1997 (62 FR

2187). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 4, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: For Dresden, Morris Area

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

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

Illinois 61021.

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: January 6, 1997

Brief description of amendments: The amendments would change the

technical specifications to clarify and maintain consistency between

the operability requirements for protective instrumentation and

associated automatic bypass features.

Date of issuance: March 14, 1997

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

Amendment Nos.: 155, 150, 174, 170

[[Page 14475]]

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 12, 1997 (62

FR 6573). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 14, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: For Dresden, Morris Area

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

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

Illinois 61021.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: November 26, 1996 as supplemented by

letters dated December 17, 1996, March 4, 1997, and March 10, 1997

Brief description of amendment: The amendment changes reactor

coolant systems pressure/temperature limits to incorporate updated

parameters and requirements.

Date of issuance: March 14, 1997

Effective date: March 14, 1997

Amendment No.: 188

Facility Operating License No. DPR-51. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4346) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 14, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Unit No. 2, Pope County, Arkansas

Date of application for amendment: April 11, 1996 as supplemented

by letters dated June 18, and September 5, 1996.

Brief description of amendment: The amendment adds low-temperature

overpressure protection requirements to the Technical Specifications as

proposed by Generic Letter 90-06.

Date of issuance: March 7, 1997

Effective date: March 7, 1997, to be implemented within 30 days.

Amendment No.: 180

Facility Operating License No. NPF-6. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20846) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 7, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: August 23, 1996, as supplemented

January 8, 1997 (TSCR 245)

Brief description of amendment: The amendment updates the pressure-

temperature limits up to 22, 27, and 32 effective full power years.

Date of Issuance: March 6, 1997

Effective date: March 6, 1997, to be implemented within 30 days of

issuance

Amendment No.: 188

Facility Operating License No. DPR-16. Amendment revises the

Technical Specifications.

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

FR 47977). The January 8, 1997, letter provided clarifying information

within the scope of the original application and did not change the

staff's initial proposed no significant hazards consideration

determination. The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated March 6, 1997 No significant

hazards consideration comments received: No.

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: October 17, 1996, as

supplemented and modified on December 13, 1996

Brief description of amendment: The amendment revises the Operating

License to reflect the transfer of Soyland Power Cooperative's 13.21-

percent minority ownership of Clinton Power Station to Illinois Power

Company. The Operating License has been revised to delete Soyland Power

Cooperative as an owner.

Date of issuance: March 13, 1997

Effective date: March 13, 1997

Amendment No.: 114

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

Operating License.

Date of initial notice in Federal Register: November 19, 1996 (61

FR 58897) and January 29, 1997 (62 FR 4337) The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

March 13, 1997. No significant hazards consideration comments received:

No

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien County, Michigan

Date of application for amendment: June 19, 1996, and supplemented

September 19, 1996, and December 20, 1996.

Brief description of amendment: The amendment revises the TS to

allow a permanent extension of the interim steam generator tube

voltage-based repair criteria for steam generator tubes used in Cycles

13, 14 and 15 at the Donald C. Cook Nuclear Power Plant, Unit 1.

Date of issuance: March 13, 1997

Effective date: March 13, 1997, with full implementation within 45

days

Amendment No.: 215

Facility Operating License No. DPR-58. Amendment revised the

Technical Specifications.

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

40022) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 13, 1997. No significant

hazards consideration comments received: No. The September 19, 1996,

and December 20, 1996, letters provided additional information within

the scope of the original application and did not change the initial

proposed no significant hazards consideration determination.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of application for amendments: May 26, 1995, and supplemented

September 26, 1995, August 2, 1996 and February 6, 1997

Brief description of amendments: The amendments revise the TS to

allow operation of Cook Unit 1 at steam generator tube plugging levels

up to 30%. Additional changes to increase operating margins for both

Unit 1 and Unit 2 are also included.

Date of issuance: March 13, 1997

Effective date: March 13, 1997, with full implementation within 45

days

[[Page 14476]]

Amendment Nos.: 214 and 199

Facility Operating License Nos. DPR-58 and DPR-74. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 19, 1995 (60 FR

37095) The September 26, 1995, August 2, 1996, and February 6, 1997,

supplements provided clarifying information that did not expand the

scope of the initial application or change the staff's proposed no

significant hazards determination. The Commission's related evaluation

of the amendments is contained in a Safety Evaluation dated March 13,

1997. No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

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 amendment revises the

Appendix A Technical Specifications relating to in-core detector

system, seismic instrumentation, meteorological instrumentation, and

turbine overspeed protection. The amendment deletes Limiting Conditions

for Operation and Surveillance Requirements related to these

instruments. The deleted requirements are to be incorporated into the

Seabrook Station Technical Requirements Manual (SSTR). The associated

Bases Sections are also deleted. In addition, Technical Specification

5.5 is deleted but will not be relocated to the SSTR. The amendment

also redesignates Paragraph 2.J of the Seabrook Operating License as

Paragraph 3, and has added new Paragraph 2.J to document the North

Atlantic commitment to relocate the above mentioned Technical

Specification requirements to the SSTR.

Date of issuance: March 12, 1997

Effective date: March 12, 1997

Amendment No.: 50

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications and Operating License.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66713). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 12, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

Connecticut

Date of application for amendment: May 23, 1996, as supplemented

July 17 and December 4, 1996

Brief description of amendment: The amendment modifies the

description of the time constants associated with the Overtemperature

Delta-T and Overpower Delta-T calculations used to establish the trip

setpoints and the time constant used in the rate-lag controller for

Steam Line Isolation, Steam Line Pressure Negative Rate-High.

Date of issuance: March 11, 1997

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

within 60 days.

Amendment No.: 134

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

Technical Specifications.

Date of initial notice in Federal Register: June 17, 1996 (61 FR

30639) The July 17 and December 4, 1996, letters provided additional,

clarifying information that did not change the scope of the May 23,

1996, application and the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated March 11, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Southern California Edison Company, et al., Docket No. 50-362, San

Onofre Nuclear Generating Station, Unit No. 3, San Diego County,

California

Date of application for amendment: February 7, 1997

Brief description of amendment: The amendment defers implementation

of Surveillance Requirement 3.1.5.4 of Technical Specification 3.1.5,

``Control Element Assembly (CEA) Alignment,'' until the next SONGS Unit

3 shutdown, which will be no later than the upcoming Cycle 9 refueling

outage (currently scheduled for April 12, 1997).

Date of issuance: March 5, 1997

Effective date: March 5, 1997

Amendment No.: 126

Facility Operating License No. NPF-15: The amendments revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes (62 FR 7477 dated February 19,

1997). The notice provided an opportunity to submit comments on the

Commission's proposed no significant hazards consideration

determination. No comments have been received. The notice also provided

for an opportunity to request a hearing by March 21, 1997, but

indicated that if the Commission makes a final no significant hazards

consideration determination any such hearing would take place after

issuance of the amendment. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated March 5, 1997.

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, California 91770

Local Public Document Room location: Main Library, University of

California, P. O. Box 19557, Irvine, California 92713

Dated at Rockville, Maryland, this 19th day of March 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

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

Reactor Regulation

[Doc. 97-7508 Filed 3-25-97; 8:45 am]

BILLING CODE 7590-01-F