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

[[Page 6567]]

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 January 17, 1997, through January 31, 1997.

The last biweekly notice was published on January 29, 1997 (62 FR

4341).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

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

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By March 14, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

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

[[Page 6568]]

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.

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

Station, Plymouth County, Massachusetts

Date of amendment request: November 26, 1996

Description of amendment request: The proposed amendment would

change the definition of ``Primary Containment Integrity,'' Note 6 on

Table 3.2.A, correct a typographical error on Table 3.2 D, correct

Table 3.2.F to reflect modifications to the plant and changes to Bases

sections 3/4.6G and 3/4.7.A. These changes are considered

administrative and have no effect on plant design, safety limit

settings or plant system operation.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed administrative changes involving typographical

errors, additions for clarity and consistency, and updating the

Bases do not affect plant design, safety limit settings, or plant

system operation and, therefore, do not modify or add any initiating

parameters that would significantly increase the probability or

consequences of any previously analyzed accident.

The changes to instrument numbers and type do not change the

parameters being surveyed or the number of operable channels for

these parameters. These changes do not modify or add any initiating

parameters and do not affect plant design, safety limit settings, or

plant system operation. Therefore, these instrument changes do not

involve a significant 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.

These proposed changes do not involve any potential initiating

events that would create any new or different kind of accident.

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 amendment does not involve a significant

reduction in a margin of safety.

These changes do not affect any safety analysis assumptions,

system operation, structures, potential initiating events or safety

limits. Therefore, it is concluded that the proposed amendment does

not involve a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Patrick D. Milano, Acting

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

Station, Plymouth County, Massachusetts

Date of amendment request: January 24, 1997

Description of amendment request: The proposed amendment will

update the Safety Limit Minimum Critical Power Ratio (SLMCPR) in

Technical Specification (TS) 2.1.2 and the associated Bases section to

reflect the results of the latest cycle-specific calculation performed

for the Pilgrim Nuclear Power Station Operating Cycle 12. In addition,

the values provided in Note 5 of Table 3.2.C.1, which are based on the

SLMCPR values, have been revised as a result of the changes to the

SLMCPR value.

Basis for proposed no significant hazards 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: 11.The proposed technical specification changes do not involve a

significant increase in the probability of an accident previously

evaluated.

The derivation of the revised SLMCPR for Pilgrim for

incorporation into the TS, and its use to determine cycle-specific

thermal limits, have been performed using NRC approved methods.

Additionally, interim implementing procedures that incorporate

cycle-specific parameters have been used which result in a more

restrictive value for SLMCPR. These calculations do not change the

method of operating the plant and have no effect on the probability

of an accident initiating event or transient.

The basis of the MCPR [minimum critical power ratio] Safety

Limit is to ensure no mechanistic fuel damage is calculated to occur

if the limit is not violated. The new SLMCPR preserves the existing

margin to transition boiling, and the probability of fuel damage is

not increased.

The basis of the MCPR criteria that define a limiting rod

pattern is to ensure the SLMCPR is not violated in the event a

control rod is fully withdrawn from the core. The new MCPR criteria

that define a limiting rod pattern continue to ensure the SLMCPR is

not violated in the event a control rod is fully withdrawn from the

core. These new criteria do not change the method of operating the

plant and have no effect on the probability of an accident

initiating event or a transient.

[[Page 6569]]

Therefore, the proposed changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes result only from a revised method of

analysis for the Cycle 12 core reload. These changes do not involve

any new method for operating the facility and do not involve any

facility modifications. No new initiating events or transients

result from these changes. Therefore, the proposed TS changes do not

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

any accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The margin of safety as defined in the TS bases will remain the

same. The new SLMCPR is calculated using NRC approved methods which

are in accordance with the current fuel design and licensing

criteria. Additionally, interim implementing procedures, which

incorporate cycle-specific parameters, have been used. The SLMCPR

remains high enough to ensure that greater than 99.9% of all fuel

rods in the core will avoid transition boiling if the limit is not

violated, thereby preserving the fuel cladding integrity.

The new MCPR criteria that define a limiting rod pattern

continue to ensure the SLMCPR is not violated in the event a control

rod is fully withdrawn from the core.

Therefore, the proposed TS changes do not involve a 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Patrick D. Milano, 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: January 29, 1997

Description of amendment request: The proposed change adds a new

entry 3.0.5 to the plant Technical Specifications (TS) to provide

specific guidance for returning equipment to service under

administrative control for the sole purpose of performing testing to

demonstrate operability.

Basis for proposed no significant hazards 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 change does not affect the operation or design of

the plant in any way. Operation of plant equipment under this change

will not differ in any way from its normal operational mode. The

normal operation of plant equipment is not a precursor to any

accident. The purpose of tests performed using this change are to

demonstrate that required automatic actions are carried out.

Equipment will be operated under administrative control for only a

short period of time. Personnel will be immediately available to

take appropriate manual action if it should be required. Therefore

operation of equipment under this change is not expected to increase

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed testing allowance does not involve any physical

alterations or additions to plant equipment or alter the manner in

which any safety-related system performs its function. 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 amendment does not involve a signifcant reduction in

the margin of safety.

Equipment will be operated under administrative control for only

a short period of time. Personnel will be immediately available to

take appropriate manual action if it should be required. The purpose

of the testing is to restore required equipment to an OPERABLE state

which increases the automatic protection available and reduces the

reliance on the compensatory measures provided by ACTION statements.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

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

Date of amendment request: October 24, 1996

Description of amendment request: The proposed amendment to the

Perry Nuclear Power Plant Technical Specifications revises those

specifications associated with the Minimum Critical Power Ratio (MCPR)

Reactor Core Safety Limit. The revision would increase the MCPR Safety

Limit values to make them more conservative.

Basis for proposed no significant hazards determination: The NRC

staff provides its analysis of the issue of no significant hazards

consideration below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

There is no change to any plant equipment, and increasing the

MCPR Safety Limit is more conservative. Therefore, the proposed

change does not significantly increase the probability or

consequences of any 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.

There are no physical changes to the plant, and increasing the MCPR

Safety Limit is more conservative. Therefore, the proposed changes do

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed MCPR Safety Limit values are more conservative, and

were calculated using NRC approved methods. Therefore, the proposed

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

The staff has reviewed the amendment request and the licensee's no

significant hazards consideration determination. Based on the review

and the above discussions, the staff proposes to determine that the

proposed changes do not involve a significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: Gail H. Marcus

[[Page 6570]]

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

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

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

Will County, Illinois

Date of amendment request: August 19, 1996

Description of amendment request: The proposed amendments revise

the steam generator (SG) repair criteria in the Byron, Unit 1, and

Braidwood, Unit 1, Technical Specifications (TS). These revisions, if

approved, would continue the use of the voltage-based SG tube repair

criteria added by Amendment No. 77, dated November 9, 1995, to the

Byron 1 TSs and by Amendment No. 69, dated November 9, 1995, to the

Braidwood 1 TSs. The subject voltage-based repair criteria are

applicable only for a specific form of SG tube degradation identified

as outer diameter stress corrosion cracking (ODSCC), which is confined

entirely within the thickness of the SG tube support plates (TSPs).

Specifically, the pending amendments for both units would continue for

one more operating cycle, the present use of a lower voltage repair

limit of 3.0 volts on the hot leg side of the SGs using the Locked-Tube

model. The cold leg side of the SGs and certain hot leg side tube/TSP

intersections (e.g., dented SG tube intersections) would continue to be

repaired using the Free-Span model. The proposed amendments are needed

because the applicability of the revised voltage-based SG tube repair

criteria for ODSCC which were added in the prior amendments cited

above, was limited to only one full operating cycle for Braidwood 1

ending in spring 1997 and for the operating cycle ending in late 1997

for Byron 1.

Additionally, the inspection and reporting requirements added to

the Byron 1 and Braidwood 1 TSs by the prior amendments cited above,

would also be continued for one more operating cycle for both units.

The maximum permissible value of the iodine-131 concentration in the

primary coolant in both the Byron 1 and Braidwood 1 TSs remains

unchanged at 0.35 microcuries per gram of coolant. Finally, the Bases

sections in the Byron 1 and Braidwood 1 TSs are proposed to be revised

to introduce the terminology associated with the Locked-Tube SG tube

model and that of the Free-Span model.

Basis for proposed no significant hazards 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.

This amendment request proposes to renew the SG tube plugging/

repair criteria previously approved by the NRC in Amendments 69 and

77 to Braidwood and Byron Technical Specifications, respectively.

The previously evaluated applicable accidents are steam

generator tube burst and Main Steam Line Break (MSLB). The

postulated MSLB outside of containment but upstream of the Main

Steam Isolation Valve (MSIV) represents the most limiting

radiological condition relative to the IPC. The potential impact on

public health and safety as a result of renewing the SG tube interim

plugging criteria contained in the current Braidwood and Byron

Technical Specifications is very low as discussed below. Tube burst

due to predominantly axially oriented ODSCC at the TSP intersections

is precluded during normal operating plant conditions since the tube

support plates are adjacent to the degraded regions of the tube in

the tube-to-tube support plate crevices.

During accident conditions, i.e., MSLB, the tubes and TSP may

move relative to each other. This can expose the crack length

portion to free-span conditions. Testing has shown that the burst

pressure correlates to the crack length that is exposed to the free-

span, regardless of the length that is still contained within the

TSP bounds.

Therefore, a more appropriate methodology has been established

for addressing leakage and burst considerations. This methodology is

based on limiting potential TSP displacements (Locked-Tube Model

Intersections) during postulated MSLB events, thus reducing the

free-span exposed crack length to minimal levels. The tube expansion

process employed in conjunction with this tube plugging criteria is

designed to provide postulated TSP displacements that result in

negligible tube burst probabilities due to the minimal free-span

exposed crack lengths. The tube expansions were performed during the

first outage that the 3.0 volt IPC was applied (Braidwood refuel

outage A1R05 -Fall 1995, and Byron midcycle outage B1P02 - Fall

1995). These expansions will be inspected in accordance with an eddy

current inspection probe that is sensitive to axial and

circumferential indications. This program will ensure the integrity

of the expansions for the additional cycle of operation. It has been

demonstrated that axial indications in the expansion region will not

result in a reduction of the load carrying capability of the

expanded tubes.

Thermal hydraulic modeling was used to determine TSP loading

during MSLB conditions. A safety factor was conservatively applied

to these loads to envelope the collective uncertainties in the

analyses. Various operating conditions were evaluated and the most

limiting operating condition was used in the analyses. Additional

models were used to verify the thermal hydraulic results.

Assessment of the tube burst probability for the Locked-Tube

Model Intersections was based on a conservative assumption that all

hot-leg TSP intersections (32,046) contained through wall cracks

equal to the postulated TSP displacement and that the crack lengths

were located within the boundaries of the TSP. Alternatively, it was

assumed that all hot-leg TSP intersections contained through wall

cracks with lengths equal to the thickness of the TSP. The

postulated TSP motion was conservatively assumed to be uniform and

equal to the maximum displacement calculated.

The total burst probability for all 32,046 through wall

indications, given a uniform MSLB TSP displacement of 0.31 inches,

was calculated to be 1x10-5. This is a factor of 1000 less than

the GL 95-05 burst probability limit of 1x10-2. Therefore, the

functional design criteria for tube expansion was to limit the TSP

motion to 0.31'' or less. However, the design goal for tube

expansion limits the TSP MSLB motion to less than 0.1''. This design

goal results in a total tube burst probability of 1x10-10 for

all 32,046 postulated through wall indications. Additional tubes

were expanded to provide redundancy for the required expansions.

The structural limit for the Locked-Tube Model Intersection SG

tube repair criteria was based on axial tensile loading requirements

to preclude axial tensile severing of the tube. Axially oriented

ODSCC does not significantly impact the axial tensile loading of the

tube. Based on the current voltage distributions and growth rates,

Monte Carlo projections were performed for Braidwood Unit 1 and

Byron Unit 1 for the additional cycle of operation that this

proposed amendment is requesting. The End of Cycle (EOC) voltage

projections for Braidwood Unit 1 Cycle 7 predict that the maximum

voltage to be seen will be less than 10.5 volts. The number of

indications predicted greater than ten volts at the end of Cycle 7

for Braidwood Unit 1 is 0.3. The EOC voltage projections for Byron

Unit 1 Cycle 9 predict that the maximum voltage to be seen will be

less than 13.5 volts. The number of indications predicted greater

than ten volts at the end of Cycle 9 for Byron Unit 1 is 4.59.

Using a tensile rupture probability for a ten volt indication of

3x10-6, the probability of tensile rupture from the predicted

0.3 indications at Braidwood is 1-(1-3x10-6)0.3 =

9.0x10-7. The probability of tensile rupture from the predicted

4.59 indications at Byron is 1-(1-3x10-6)4.59 =

1.38x10-5. Both of these probabilities result in a negligible

contribution to the total burst probability when compared to the

1x10-2 GL 95-05 limit.

Cellular corrosion is a more limiting mode of degradation at the

TSPs with respect to affecting the tube structural limit. Tensile

tests that measure the force required to sever a tube with cellular

corrosion and uncorroded cross sectional areas are used to establish

the lower bound structural limit. Based upon these tests, a lower

bound 95% confidence level structural voltage limit of 37 volts was

established for cellular corrosion. This limit meets the Regulatory

Guide (RG) 1.121, ``Basis for Plugging Steam Generator Tubes,''

structural requirements based upon the normal operating pressure

differential with a safety factor of 3.0 applied. Due to the limited

database supporting this value, the

[[Page 6571]]

structural limit was conservatively reduced to 20 volts. Accounting

for voltage growth and Non-Destructive Examination (NDE)

uncertainty, the full IPC upper limit exceeds ten volts. However,

for added conservatism a single voltage repair limit of 3.0 volts

for the Locked-Tube Model Intersection indications is specified in

the current plugging/repair criteria. All indications at the Locked

Tube Model Intersections with bobbin coil probe voltages greater

than 3.0 volts will be plugged or repaired.

The free-span tube burst probability must be calculated for the

indications at the Free-Span Model Intersections. The total burst

probability must be within the requirements of GL 95-05. The free-

span structural voltage limit is calculated using correlations from

the database described in GL 95-05, with the inclusion of the recent

Byron, Braidwood, and South Texas tube pull results. The structural

limit for the Free-Span Model Intersections is 4.745 volts. The

lower voltage repair limit for the indications at the Free-Span

Model Intersections continues to be 1.0 volt. The upper voltage

repair limit for the indications at the Free-Span Model

Intersections will be calculated in accordance with GL 95-05.

Since IPC will not be applied to indications at the Flow

Distribution Baffle (FDB), no leakage or burst analyses are required

for these indications.

Per GL 95-05, MSLB leak rate and tube burst probability analyses

are required to be performed prior to returning the unit to power.

The results of these analyses are to be included in a report to the

NRC within 90 days of restart. If allowable limits on leak rates and

burst probability are exceeded, the results are to be reported to

the NRC and a safety assessment of the significance of the results

is to be performed prior to returning the SGs to service.

A site specific calculation has determined the site allowable

leakage limit for Braidwood and Byron. These limits use the

recommended Dose Equivalent Iodine-131 transient spiking values

consistent with NUREG-0800, ``Standard Review Plan'' and ensure site

boundary doses are within a small fraction of the 10 CFR 100

requirements.

The projected leakage rate calculation methodology described in

WCAP-14046, ``Braidwood Unit 1 Technical Support for Cycle 5 Steam

Generator Interim Plugging Criteria,'' and WCAP-14277, ``SLB Leak

Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP

Intersections,'' will be used to calculate the EOC leakage. This

method includes a Probability of Detection (POD) value of 0.6 for

all voltage amplitude ranges and uses the accepted leak rate versus

bobbin voltage correlation methodology (full Monte Carlo) for

calculating the leak rate, as described in GL 95-05. The database

used for the leak and burst correlations is consistent with that

described in GL 95-05 with the inclusion of the Byron Unit 1,

Braidwood Unit 1, and South Texas tube pull results. The EOC voltage

distribution is developed from the POD adjusted beginning-of-cycle

(BOC) voltage distributions and uses Monte Carlo techniques to

account for variances in growth and NDE uncertainty.

The Electric Power Research Institute (EPRI) leak rate

correlation has been used. This correlation is based on free-span

indications that have burst pressures above the MSLB pressure

differential. There is a low but finite probability that indications

may burst at a pressure less than MSLB pressure. With limited TSP

motion for the Locked-Tube Model Intersections, the tube is

constrained by the TSP and tube burst is precluded. However, the

flanks of the crack may open up to contact the Inside Diameter (ID)

of the TSP hole and result in a primary-to-secondary leak rate

potentially exceeding that obtained from the EPRI correlation. This

phenomenon is known as an Indication Restricted from Burst (IRB)

condition.

ComEd has performed laboratory testing to determine the bounding

leak rate obtainable in an IRB condition (6.0 gallons per minute).

The bounding leak rate value was then applied to a leak rate

calculation methodology that accounts for the MSLB leak rate

contribution from IRB indications to the total leak rate calculated

as described above. Results indicate that the IRB contribution to

the total leak rate value is negligible. However, ComEd will

conservatively add a leakage contribution due to IRBs in addition to

the leakage calculated in accordance with GL 95-05. When this is

done, the dose at the site boundary resulting from the predicted

leakage will be a small fraction (less than 10%) of the 10 CFR 100

limits.

Modification of the Braidwood and Byron TS to clarify

application of the proposed tube plugging/repair criteria is purely

administrative and will not have any effect on the probability or

consequences of an accident previously evaluated.

Operating experience over the last cycle with this plugging

criteria applied has not revealed any unpredicted or unusual

effects.

For these reasons, renewal of the current Braidwood and Byron

tube plugging criteria does not adversely affect SG tube integrity

and results in acceptable dose consequences. By effectively

eliminating tube burst at the Locked-Tube Model TSP intersections,

the likelihood of a tube rupture is substantially reduced and the

probability of occurrence of an accident previously evaluated is

reduced.

This conclusion is not affected by foreign or domestic plant SG

experiences (NRC Information Notice 96-09 and its supplement). As

the following evaluation shows, these experiences are not relevant

to Braidwood or Byron.

A foreign unit detected eddy current signal distortions in one

area of the top TSP during a 1995 inspection. The steam generators

had been chemically cleaned in 1992. Visual inspection showed that a

small section of the top TSP had broken free and was resting next to

the steam generator tube bundle wrapper. The support plate showed

indications of metal loss.

The chemical cleaning process used by the foreign unit was

developed by the utility and differs significantly from the modified

EPRI/SGOG process performed at Byron Unit 1 in 1994. The foreign

chemical cleaning process, coupled with the specific application of

the process, resulted in TSP corrosion of up to 250 mils compared to

a maximum of 2.16 mils (11 mils maximum allowed) measured at Byron.

During the Byron eddy current inspection performed after the

chemical cleaning, no distortion of the tube support plate signals

was reported. Therefore, these differences in cleaning processes

imply that this foreign experience is irrelevant to the effects of

the chemical cleaning process on the TSPs at Byron. Chemical

cleaning of the SGs has not occurred at Braidwood.

A number of units have experienced TSP cracking associated with

severe tube denting due to TSP corrosion at the tube-to-TSP crevice.

WCAP-14273, Section 12.4, shows that a diametral reduction of a SG

tube of 0.065 inches is required to develop stress levels above

yield in the TSP ligaments at dented intersections. The bobbin

voltage range associated with a one mil radial dent is twenty to

twenty-five volts.

Although Braidwood Unit 1 and Byron Unit 1 have not seen

corrosion induced denting, a 0.610 inch diameter bobbin coil probe

will be used as a go/no-go gauge to assess dents at the Locked-Tube

Model Intersections, if they occur in the future. If a tube has a

dent at a Locked-Tube Model TSP intersection that fails to pass the

go/no-go test probe, IPC will not be applied to that intersection.

In addition, if the dent is determined to be corrosion induced, the

Free-Span Model repair criteria will be applied to the intersections

adjacent to the dented intersection. IPC repair limits will not be

applied to tubes with dents greater than 5.0 volts since dent

signals of this magnitude could mask a 1.0 volt ODSCC signal. Tube

intersections with corrosion induced dents greater than 5.0 volts

and the intersections adjacent to such an intersection were not

selected for tube expansion to preclude adverse effects of the

failure of such a tube on limiting TSP displacement. If corrosion

induced denting, either greater than 5.0 volts or such that the tube

is unable to pass a 0.610 inch diameter bobbin coil probe, are

detected at an intersection adjacent to an expanded intersection,

the dented intersection will be inspected by an EPRI developed

technique to determine if the TSP is cracked. If a crack-like

indication is identified in a TSP, a plus point inspection will be

conducted per the EPRI TSP program. If the plus point inspection

verifies the existence of a crack-like indication, the effect of

that indication on TSP displacement will be evaluated. If this

evaluation shows that TSP displacement would be greater than 0.1

inches during a MSLB event, the effected area will either be

mechanically corrected or the Free-Span Model criteria will be

applied to the affected area. Based on the information presented

above, the SG tube denting experience at other plants is not

relevant to Braidwood or Byron.

A foreign utility's SGs have experienced cracking at the top

TSP. The cause of the cracking appears to be the configuration of

the single anti-rotation device, connected between the SG shell and

wrapper, and the wrapper internals. The single anti-rotation device

carries the full load associated with the wrapper to shell motion.

This rotational load is believed to be transferred to the TSP via

the wrapper internals. The Byron/Braidwood Unit 1 SG design (D-4)

uses three anti-rotation devices to spread the rotational load. The

D-4 wrapper internals are configured such that this load is not

directly transmitted to the TSP.

[[Page 6572]]

No top TSP cracking has been detected at Braidwood Unit 1 or

Byron Unit 1 and very few (-10 with all tube expansions

functional. The burst probability for the Free-Span Model TSP

intersections will be dependent on the number and size of

indications at these applicable intersections. The total burst

probability will be within the limit specified in GL 95-05.

Accounting for the unlikely event of a failure of the expanded

tubes, a sufficient number of redundant expansions exist to ensure

that the burst probability remains below 1x10-5. This includes

the conservative assumption that all 32,046 hot-leg TSP

intersections contain through wall indications. This level of burst

probability is considered to be negligible when compared to the GL

95-05 limit of 1x10-2.

In addressing the combined effects of a Loss Of Coolant Accident

(LOCA) during a Safe Shutdown Earthquake (SSE) on the SG as required

by General Design Criteria (GDC) 2, it has been determined that tube

collapse may occur in the steam generators at some plants. The tube

support plates may become deformed as a result of lateral loads at

the wedge supports located at the periphery of the plate due to the

combined effects of the LOCA rarefaction wave and SSE loadings. The

resulting pressure differential on the deformed tubes may cause some

of the tubes to collapse. There are two issues associated with SG

tube collapse. First, the collapse of SG tubing reduces the Reactor

Coolant System (RCS) flow area through the tubes. The reduction in

flow area increases the resistance to flow of steam from the core

during a LOCA which, in turn, may potentially increase the Peak Clad

Temperature (PCT). Second, there is a potential that partial through

wall cracks in the SG tubes could progress to through wall cracks

during tube deformation or collapse. The tubes subject to collapse

have been identified via a plant specific analysis and are excluded

from application of any voltage-based criteria. This analysis is

included in revision 3 to WCAP-14046 which was submitted to the NRC

June 19, 1995.

Modification of the Braidwood and Byron Technical Specifications

to clarify application of the proposed tube plugging/repair criteria

is purely administrative and will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Operating experience over the last cycle with this plugging

criteria applied has not revealed any unpredicted or unusual

effects.

SG tube integrity will continue to be maintained following

renewal of the 3.0 volt IPC voltage repair limit through inservice

inspection, tube repair and primary-to-secondary leakage monitoring.

By effectively eliminating tube burst at the Locked-Tube Model TSP

Intersections, the potential for multiple tube ruptures is

essentially eliminated.

ComEd has evaluated industry experiences with TSP degradation,

eddy current signal distortions, and component misalignment. Eddy

current signal distortions due to TSP geometry are not indicative of

TSP degradation and do not result in any kind of new or different

accident.

The component misalignment experienced by one unit is not

applicable to Braidwood Unit 1 or Byron Unit 1 and, thus, will not

result in any kind of new or different accident. Specific

limitations, as discussed in response to Question 1, will be applied

to indications at the Locked-Tube Model Intersections which contain

dents. These limitations ensure that the integrity of the SG tubes

is maintained consistent with the current analyses should tube

denting or TSP cracking occur.

Therefore, renewal of the current tube plugging/repair criteria

at Braidwood Unit 1 and Byron Unit 1 will not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The use of the voltage-based, bobbin coil, tube support plate

plugging criteria with tube expansion at Braidwood Unit 1 and Byron

Unit 1 is demonstrated to maintain SG tube integrity commensurate

with the criteria of RG 1.121. RG 1.121 describes a method

[[Page 6573]]

acceptable to the NRC staff for meeting GDC 14, 15, 31, and 32 by

reducing the probability or the consequences of steam generator tube

rupture.

Reducing the probability or the consequences of steam generator

tube rupture is accomplished by determining an eddy current

inspection voltage value which represents a limit for leaving an

axial, crack-like indication at an in service SG tube TSP

intersection. Tubes with ODSCC voltage indications beyond this

limiting value must be removed from service by plugging or repaired

by sleeving. Implementation of a 3.0 volt IPC voltage repair limit

for the Locked-Tube Model Intersections has been evaluated and shown

not to present a credible potential for a steam generator tube

rupture event during normal or faulted plant conditions, even with

worst case assumptions. The total tube burst probability will

include a contribution from the indications at the Locked-Tube Model

Intersections and from indications at the Free-Span Model

Intersections. The projected EOC voltage distribution of crack-like

indications at the TSP elevations will be confirmed to result in

acceptable primary-to-secondary leakage during all plant conditions

such that radiological consequences are not adversely impacted.

Addressing RG 1.83 considerations, implementation of the

increased Locked-Tube Model Intersection bobbin coil voltage-based

repair criteria is supplemented by enhanced eddy current inspection

guidelines to provide consistency in voltage normalization and a

100% eddy current inspection sample size at the affected TSP

elevations.

For the leak and burst assessments, the population of

indications in the EOC voltage distribution is dependent on the POD

function. The purpose of the POD function is to account for new

indications that may develop over the cycle, and to account for

indications not identified by the data analyst. In implementing this

proposed IPC renewal, ComEd will continue to use the conservative GL

95-05 POD value of 0.6 for all voltage amplitude ranges.

Modification of the Braidwood and Byron Technical Specifications

to clarify application of the proposed tube plugging/repair criteria

is purely administrative and will not reduce any safety margins.

Operating experience over the last cycle with this plugging

criteria applied has not revealed any unpredicted or unusual

effects.

Implementation of the TSP elevation repair limits will decrease

the number of tubes which must be repaired. Installation of steam

generator tube plugs or sleeves reduces the RCS flow margin. Thus,

implementation of the IPC will maintain the margin of flow that

would otherwise be reduced in the event of increased tube plugging

or sleeving.

As discussed previously, ComEd has evaluated industry

experiences with TSP degradation, eddy current signal distortions,

and component misalignment. Eddy current signal distortions at tube

support plates will be evaluated to attempt to determine the cause

of the distortion. A signal distortion alone will not result in

reduction in the margin of safety. The foreign unit that experienced

the component misalignment was of a significantly different design

than the Braidwood Unit 1 and Byron Unit 1 steam generators.

Analysis of the design differences shows that component misalignment

of that type is not applicable to Braidwood Unit 1 or Byron Unit 1

and, thus, will not result in a reduction in the margin of safety.

An inspection was conducted during the last Braidwood Unit 1 and

Byron Unit 1 refueling outages which verified this conclusion.

Specific limitations, as discussed previously, will be applied

to indications at the Locked-Tube Model Intersections which contain

dents. These limitations conservatively treat indications as free-

span to ensure that the integrity of the SG tubes is maintained

consistent with current analyses should tube denting or TSP cracking

occur. Application of the 3.0 volt Locked-Tube Model Intersection

IPC and the 1.0 volt Free-Span Model Intersection IPC at Braidwood

Unit 1 and Byron Unit 1, with the limitations specified, will not

result in a reduction in a margin of safety.

Thus, the implementation of this amendment does not result in a

significant reduction in a margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

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

Wilmington, Illinois 60481.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendment request: January 6, 1997

Description of amendment request: The proposed amendment would

clarify and maintain consistency between the operability requirements

for protective instrumentation and associated automatic bypass

features.

Basis for proposed no significant hazards 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 are administrative in nature and do not

affect the probability or consequences of any previously evaluated

accidents for Dresden or Quad Cities Stations. The proposed

amendment is consistent with the current safety analyses and

represents sufficient requirements for the continued assurance and

reliability of the RPS and Rod Block Instrumentation equipment,

which is assumed to operate in the safety analysis, or provides

continued assurance that specified parameters associated with RPS

and Rod Block Instrumentation remain within their acceptance limits.

Therefore, these changes will not affect the probability or

consequences of a previously evaluated accident.

The RPS and Rod Block Instrumentation related to this proposed

amendment is not assumed in any safety analysis to initiate any

accident sequence for Dresden or Quad Cities Stations; therefore,

the probability of any accident previously evaluated is not affected

by the proposed amendment.

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

from any accident previously evaluated because:

The proposed changes are administrative in nature and serve to

maintain consistent and clear requirements for operability as

specified in the Technical Specifications for the Limiting

Conditions for Operation and Surveillance Requirements for the RPS

and Rod Block Instrumentation. No new modes of operation or changes

to any plant equipment are proposed by the proposed amendment

request. The associated systems related to this proposed amendment

are not assumed in any safety analysis to initiate any accident

sequence for Dresden or Quad Cities. The proposed changes maintain

the present level of operability; and therefore, the proposed

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

accident than any previously evaluated.

3) Involve a significant reduction in the margin of safety

because:

The proposed changes are administrative in nature and do not

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

Dresden and at Quad Cities based on RPS and Rod Block

Instrumentation system design, safety analysis requirements and

operational performance. Since the proposed changes are

administrative in nature and maintain necessary levels of the RPS

and Rod Block reliability, the proposed changes do not involve a

reduction in the margin of safety.

The proposed amendment for Dresden and Quad Cities Stations will

not reduce the availability of the RPS and Rod Block Instrumentation

System which is required to mitigate accident conditions; therefore,

the proposed changes do not involve a reduction in the margin of

safety.

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

this

[[Page 6574]]

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Area

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

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

Illinois 61021

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: October 22, 1996

Description of amendment request: The proposed amendments would

allow continued plant operation at elevated Containment Lower

Compartment temperatures between 125 deg. and 135 deg. F for a period

not to exceed 72 cumulative hours.

Basis for proposed no significant hazards 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 amendment involve a significant increase in

the probability or consequences of any accident previously evaluated

in the UFSAR [Updated Final Safety Analysis Report]?

The increase in maximum Containment Lower Compartment

temperature will not change the operation of any equipment which is

important to safety. All components and instruments will continue to

perform as designed in the higher temperature environment for the

period that the revised Technical Specification allows. This

temperature increase will not impact the ability of any component or

instrument to perform its function in the event of an accident.

Therefore, the probability of an accident is not impacted. The

increased temperature will cause a decrease in the air mass in lower

containment. This change has been evaluated for impact on

containment temperature and pressure in accident conditions. The air

mass change is conservative for peak containment pressure since the

air mass is decreased. Maximum containment temperatures during a

postulated accident are slightly increased as a result of higher

initial Containment Lower Compartment temperature. The increase in

peak temperature remains within the allowable values and thus does

not increase the probability or consequence of an accident. The

minimum containment pressure as a result of steam condensation in

containment is lowered as a result of the decreased air mass in

containment. Due to the conservative assumptions made in modeling

containment for minimum pressure response, this change has no impact

on the accident analysis.

Based on the analysis of the bounding accidents that may be

impacted by increased Containment Lower Compartment temperature and

the review of the effect of the increased temperature on components

in lower containment, it is determined that the probability and

consequence of any analyzed accident is unchanged as a result of

this change.

2. Does the proposed amendment create the possibility of a new

or different kind of accident not previously evaluated?

The revised maximum Containment Lower Compartment temperature

will not change any systems or operations procedures except to

procedurally respond should Containment Lower Compartment

temperature remain elevated for a period near the revised limiting

period. The response of the systems and components are unaffected by

this change. All instruments are qualified for the revised service

conditions and will perform in the same manner as before. Normal

operation and transient response will remain unchanged. Review of

previously analyzed accidents show that no new transients are

created as a result of this change. Based on this review there are

no new or different accidents made possible by this change.

3. Does the proposed amendment involve a significant reduction

in a margin of safety?

The amendment could potentially affect the containment system.

The operation and analysis of the reactor coolant system and fuel

are unaffected by this change. The maximum containment temperature

is slightly increased while the maximum containment pressure is

decreased. The minimum containment pressure could be slightly

decreased and minimum containment temperature is unaffected. All

these parameters have been reviewed and determined to be within

assumptions made in these analyses. The accident transient analyses

are unaffected beyond these small changes and remains acceptable in

all cases. Therefore, the margin of safety is unaffected by this

amendment.

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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: January 6, 1997

Description of amendment request: The proposed amendments would

allow a one-time revision to Technical Specifications 3.6.1.1, 3.6.1.2,

3.6.1.8, and 3.6.1.9 to allow operation of the Containment Purge

Ventilation System (VP) during Modes 3 and 4 following the steam

generator (SG) replacement outage. This one-time revision would be

necessary due to respiratory hazardous gases released during heatup

after the replacement of the SGs. The VP system would be used to remove

the hazardous gases.

Basis for proposed no significant hazards 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 activity does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The VP [Containment Purge Ventilation] System has no interfaces

with any primary system, secondary system, or power transmission

system. It has no interfaces with any reservoir of radioactive gases

or liquids. None of the systems listed above are modified by the

activity. In summary, no ``accident initiator'' is affected with the

proposed operation of the VP System in Modes 3 and 4. For this

reason, the activity does not involve an increase in the probability

of an accident previously evaluated.

Analyses have been performed to determine upper bounds to the

source term, the offsite doses, and the Control Room dose. The

results of that analyses are reported above. Both the source term

and the doses were found to be significantly lower than the results

of the corresponding design basis analyses. In addition, it has been

determined that with no credit taken for any heat transfer from the

fuel and cladding to the moderator channels, that sufficient time

would exist for the operators to initiate recovery of flow from the

ECCS [Emergency Core Cooling System] to the reactor core. The flow

required from the ECCS to maintain the core in a coolable geometry

was found to be well within the capacity of any one ECCS pump.

Furthermore, it was determined that convective heat transfer to

steam would be sufficient to prevent release of significant source

term or a significant degree of fuel damage.

For the above reasons, it is determined that operation of the VP

System in Mode 3 or 4 immediately following the steam generator

replacement outage does not involve a significant increase in either

the probability or the consequences of an accident previously

evaluated.

2. The activity does not create the possibility of a new or

different type of

[[Page 6575]]

accident from any accident previously evaluated.

As discussed above, no ``accident initiators'' are affected by

the proposed activity. Operation of the VP System proposed for Modes

3 and 4 will be the same as that routinely carried in other modes of

operation. For these reasons, the activity will not create the

possibility of a new or different type of accident from any

previously evaluated.

3. The activity does not involve a significant reduction in the

margin of safety.

Margin of safety is associated with confidence in the ability of

the fission product barriers (the fuel and fuel cladding, the

Reactor Coolant System pressure boundary, and the containment) to

limit the level of radiation doses to the public. The proposed

operation of the VP System will occur at the end of an extended

outage. The level of decay heat and activity in the reactor is very

low compared to the level of decay heat and activity associated with

full power operations. For this reason, the likelihood of damage to

the fuel following a DBLOCA [design basis loss-of-coolant accident]

occurring during the proposed purging is reduced, as determined

above. Both offsite doses and doses to the Control Room were found

to be small compared to the limits of 10 CFR [Part] 100 and GDC

[General Design Criterion] 19. For these reasons, the activity 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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: January 13, 1997

Description of amendment request: The proposed amendments would

implement the performance-based containment leak rate testing

requirements of 10 CFR Part 50, Appendix J, Option B, for Type A

testing.

Compliance with 10 CFR Part 50, Appendix J, provides assurance that

the primary containment, including those systems and components that

penetrate the primary containment, do not exceed the allowable leakage

rate values specified in the Technical Specifications and Bases. The

allowable leakage rate is determined so that the leakage assumed in the

safety analyses is not exceeded.

On February 4, 1992, the NRC published a notice in the Federal

Register (57 FR 4166) discussing a planned initiative to begin

eliminating requirements marginal to safety that impose a significant

regulatory burden. Appendix J to 10 CFR Part 50, ``Primary Containment

Leakage Testing for Water-Cooled Power Reactors,'' was considered for

this initiative and the staff undertook a study of possible changes to

this regulation. The study examined the previous performance history of

domestic containments and examined the effect on risk of a revision to

the requirements of Appendix J. The results of this study are reported

in NUREG-1493, ``Performance-Based Leak-Test Program.''

Based on the results of this study, the staff developed a

performance based approach to containment leakage rate testing. On

September 12, 1995, the NRC approved issuance of this revision to 10

CFR Part 50, Appendix J, which was subsequently published in the

Federal Register on September 26, 1995, and became effective on October

26, 1995. The revision added Option B ``Performance-Based

Requirements'' to Appendix J to allow licensees to voluntarily replace

the prescriptive testing requirements of Appendix J with testing

requirements based on both overall and individual component leakage

rate performance.

Regulatory Guide 1.163, ``Performance-Based Containment Leak Test

Program,'' was developed as a method acceptable to the staff for

implementing Option B. Accordingly, the licensee has submitted, in its

application dated January 13, 1997, proposed changes to the TS to

implement 10 CFR Part 50, Appendix J, Option B, by referring to

Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test

Program.''

Basis for proposed no significant hazards 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 will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Containment leak rate testing is not an initiator of any

accident; the proposed change does not affect reactor operations or

accident analysis, and has no significant radiological consequences.

... Therefore, this proposed change will not involve an increase in

the probability or consequences of any previously-evaluated

accident.

2. The proposed change will not create the possibility of any

new accident not previously evaluated.

The proposed change does not affect normal plant operations or

configuration, nor does it affect leak rate test methods. The test

history at McGuire (two consecutive successful tests) provides

continued assurance of the leak tightness of the containment

structure.

3. There is no significant reduction in a margin of safety.

The proposed changes are based on NRC-accepted provisions, and

maintain necessary levels of reliability of containment integrity.

The performance-based approach to leakage rate testing recognizes

that historically good results of containment testing provide

appropriate assurance of future containment integrity; this supports

the conclusion that the impact on the health and safety of the

public as a result of extended test intervals is negligible. In

addition, local leak[]rate testing will continue to provide

assurances of overall containment integrity.

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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: October 16, 1996

Description of amendment request: The amendment requests to change

the Waterford 3 Technical Specifications Table 4.3-1 to expand the

applicability for Core Protection Calculator operability and to allow

for the application of a Cycle Independent Shape Annealing Matrix.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

The proposed change will reduce the amount of non-conservatism

presently allowed for linear power level, the CPC delta T power, and

CPC nuclear power signals.

[[Page 6576]]

Changing the tolerance range from plus or minus 2% to between -0.5%

and 10% between 15% and 80% RATED THERMAL POWER, except during

physics testing, will allow more conservative settings than

currently allowed. The consequences of an accident will be reduced

due to the proposed change because it is less likely to be non-

conservative in power.

This proposed change will allow use of Cycle Independent Shape

Annealing Matrix (CISAM) elements. These elements will be validated,

during startup testing, by monitoring the same parameters used for

cycle specific shape annealing matrix (SAM) elements. If the CISAM

is determined to be no longer valid, a cycle specific SAM will be

calculated and used in the CPC's. In addition, use of CISAM gives

better agreement throughout the cycle.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed change to TS power calibration tolerance limits is

conservative relative to the current TS requirement. CPC's cannot

cause an accident and this change will not create the possibility of

a new or different type accident. The changes ensures that the

reactor will trip prior to the current condition due to higher CPC

power.

As stated previously, CISAM modeling removes some of the

uncertainty associated with axial shape and provides increased

assurance that the CPC is appropriately modeling the core.

Therefore, the proposed changes will not create the possibility

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

evaluated.

The proposed change to the TS reduces the amount of non-

conservatism in safety system power indications and maintains the

margin of safety for design basis events which take credit for the

linear power level, the CPC delta T power, and CPC nuclear power

signals.

CISAM will be validated each cycle during startup testing and

must meet the same parameters as cycle specific SAM elements. Since

CISAM has a better accuracy than the cycle dependent SAM, the margin

of safety is improved.

Therefore, the proposed changes 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

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: November 27, 1996

Description of amendment request: The proposed change request would

change the acceptance criteria for the individual cell voltage from

2.0v to 2.09v, change the surveillance frequency for battery specific

gravities to implement the recommendations of IEEE 450-1995, delete

surveillance requirement 4.7.B.4.d, add a clarifying phrase ``while on

a float charge....'' where appropriate, and update the Basis to reflect

these changes.

Basis for proposed no significant hazards 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:

This request has been determined to involve No Significant

Hazards in that it does not:

1. Involve a significant increase in the probability or

consequences

of an accident previous[ly] evaluated; (or)

The proposed change in ICVs [individual cell voltages] does not

increase the probability of an accident previously evaluated, as it

increases the required voltage for each ICV.

The proposed change in frequency does not increase the

probability or consequences of an accident previously evaluated, as

the change in the frequency of specific gravity testing is the

result of industry experience gained over the years. The weekly

reading of pilot cell specific gravity and cell voltage, along with

the quarterly reading of all ICVs and a 10% sample of specific

gravities from designated cells provides an acceptable means of

determining cell operability as specified in IEEE 450-1995.

The proposed deletion of Technical Specification Surveillance

Requirement 4.7.B.4.d only removes an unnecessary Technical

Specification surveillance and is consistent with NUREG-1433,

Standard Technical Specifications General Electric Plants, BWR/4,

Revision 1, April 1995. No change to plant systems, components or

operating conditions are associated with this change. Existing

Technical Specification station and diesel generator battery

inspection and testing requirements adequately verify battery

operability and condition.

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

from any accident previous[ly] evaluated; (or)

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

different kind of accident than previously evaluated, as the change

only involves raising a required voltage, performing an existing

surveillance on a different frequency, and removing an unnecessary

annunciator surveillance requirement. The station battery and diesel

generator battery low voltage annunciator setpoints do not meet any

of the criteria codified in 10 CFR 50.36 for determining content of

Technical Specifications and removal of surveillance requirement is

consistent with NUREG-1433, Standard Technical Specifications

General Electric Plants, BWR/4, Revision 1, April 1995. There is no

change to hardware or operating conditions.

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

The proposed change to the ICV does not decrease the margin of

safety, as increasing the required voltage actually increases the

margin of safety. The proposed change to the frequency does not

decrease the margin of safety as it continues to require testing and

evaluation of the requisite surveillance points and implements

requirements which have been determined to provide an adequate level

of safety by the IEEE. The removal of Technical Specification

surveillance requirements for the battery low voltage annunciator

setpoints does not affect any plant systems, components or operating

conditions.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: Patrick D. Milano, Acting

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: January 23, 1997, as revised by letter

dated January 28, 1997.

Description of amendment request: The proposed amendment would make

changes to Section 3.5/4.5.C of the technical specification (TS) bases

to clarify the minimum residual heat removal (RHR) and residual heat

removal service water (RHRSW) pump requirements for post-accident

containment heat removal. In conjunction with the proposed amendment,

the licensee requested NRC staff review and approval of an update to

the design basis accident containment temperature and pressure response

for the limiting single failure (loss of diesel generator) which

results in minimum RHR and RHRSW pump availability.

Basis for proposed no significant hazards determination: As

required by

[[Page 6577]]

10 CFR 50.91(a), the licensee has provided its analysis of the issue of

no significant hazards consideration. The NRC staff has reviewed the

licensee's analysis against the standards of 10 CFR 50.92(c). The NRC

staff's review is presented below.

(1) The proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment will change TS bases to clarify the

minimum RHR and RHRSW pump requirements for post-accident

containment heat removal. The proposed amendment will also correct

an error in a previous analysis on containment temperature and

pressure response following a design basis accident (DBA) that was

submitted for the NRC staff review on May 1, 1986. The proposed

amendment does not affect the physical configuration of the plant or

how it is operated. The licensee's analysis, using a new decay heat

model, determined that the calculated maximum suppression pool

temperature will be 2 degrees Fahrenheit greater (184 degrees

Fahrenheit vs. 182 degrees Fahrenheit) than that predicted in its

previous analysis, based on an earlier decay heat model, that was

submitted for the NRC staff review on May 1, 1986. The licensee

evaluated the effects of this increase on emergency core cooling

system (ECCS) pump net positive suction head, wetwell attached

piping, and environmental conditions in the ECCS pump rooms, and

concluded that the change is acceptable. The consequences or

probability of a previously evaluated accident will, therefore, not

be significantly increased.

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

or different kind of accident from any accident previously

evaluated.

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

different kind of accident from any previously evaluated since the

proposed amendment does not affect the physical configuration of the

plant or how it is operated. The proposed amendment revises the TS

bases to clarify the minimum RHR and RHRSW pump requirements for post-

accident containment heat removal.

(3) The proposed changes do not result in a significant

reduction inthe margin of safety.

The proposed amendment will change TS bases to clarify the

minimum RHR and RHRSW pump requirements for post-accident

containment heat removal. The proposed amendment will also correct

an error in a previous analysis on containment temperature and

pressure response following a design basis accident (DBA) that was

submitted for the NRC staff review on May 1, 1986. The proposed

amendment does not affect the physical configuration of the plant or

how it is operated. The licensee's analysis, using a new decay heat

model, determined that the calculated maximum suppression pool

temperature will be 2 degrees Fahrenheit greater (184 degrees

Fahrenheit vs. 182 degrees Fahrenheit) than that predicted in its

previous analysis, based on an ealier decay heat model, that was

submitted for the NRC staff review on May 1, 1986. The licensee

evaluated the effects of this increase on emergency core cooling

system (ECCS) pump net positive suction head, wetwell attached

piping, and environmental conditions in the ECCS pump rooms, and

concluded that the change is acceptable. Therefore, the proposed

amendment does not involve a significant reduction in a margin of

safety.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Gerald Charnoff, 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: December 9, 1996

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 the reactor trip system

(RTS) and engineering safety features actuation systems (ESFAS)

instrumentation channels, and make certain changes in trip setpoints

and allowance values due to a setpoint methodology change in support of

the calibration extensions. Channel operational tests (COTs) and trip

actuating device operational tests (TADOTs) associated with these

channels are also being extended. Revisions to the appropriate TS Bases

are being revised to support the TS revisions.

Basis for proposed no significant hazards 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 channel calibration, COT, and TADOT interval

increases from 18 to 24 months, the setpoint change, and the

allowable value changes do not alter the intent or method by which

the channel calibrations 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 calibration and

maintenance histories indicate that the equipment will continue to

perform satisfactorily with longer surveillance intervals. With the

exception of the pressurizer water level - high instrument, no

recurring surveillance or maintenance problems were identified for

the RTS or ESFAS instrumentation channels.

The pressurizer water level instruments do not have a safety

limit and are not credited in the DCPP safety analysis. The

recurring surveillance problems were mainly due to calibration zero

shift which is reflected in the statistically determined drift and

in the proposed pressurizer water level high setpoint. The zero

shift problem of these transmitters was a recurring problem with the

calibration procedure. The procedures for calibrating these

instruments have been revised to improve the repeatability of the

surveillance activity.

The trip setpoint and allowable value changes for pressurizer

water level - high are each in the more restrictive direction. The

revised setpoint would tend to trip the reactor sooner than the

present settings. These changes ensure that sufficient margin is

maintained for the pressurizer water level to accommodate the

channel statistical uncertainty resulting from a 30-month operating

cycle.

A statistical analysis of channel uncertainty for a bounding 30-

month operating cycle has been performed. There is sufficient margin

between the existing TS limits and the licensing basis safety

analysis limits to accommodate the channel statistical uncertainty

resulting from a 30-month operating cycle. The existing margin

between the TS limits and the safety analysis limits provides

assurance that plant protective actions will occur as required.

However, a change to the safety analysis limit is proposed in order

to provide additional margin for the RCS loss of [f]low-low

setpoint.

Westinghouse has evaluated the safety analysis limit for the RCS

loss of flow-low setpoint and has determined that the limit can be

changed from 87 percent of MMF to 85 percent of MMF with no impact

on the probability and insignificant impact on the consequences of

accidents already analyzed. The existing conclusions of the DCPP

FSAR Update remain valid with the safety analysis limit change.

Using the new safety analysis limit, sufficient margin exists

between the TS limit and the safety analysis limit to accommodate

the channel statistical uncertainty resulting from a 30-month

operating cycle.

The proposed changes to the allowable values ensure that drift

assumptions regarding the protection racks and direct input

functions are met.

There are no known mechanisms that would significantly degrade

the performance of the evaluated instrument channels during normal

plant operation. All potential time-

[[Page 6578]]

related degradation mechanisms have insignificant effects in the

time frame of interest (maximum of 30 months). PG&E will continue to

perform the maintenance required to maintain the qualification of

this safety related equipment.

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.

The proposed pressurizer water level trip setpoint, RCS flow

safety analysis limit, and various allowable value changes provide

adequate margin to accommodate instrument channel uncertainty over a

30-month operating cycle. Plant equipment, which will be set at, or

more conservative than, the trip setpoints, will provide protective

functions to assure that the safety analysis limits are not

exceeded. The change to the RCS loss of flow safety analysis limit

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

accident since the setpoint will remain as currently specified and

only results in an insignificant delay in the plant response to the

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 change does not involve a significant reduction

in a margin of safety.

For almost all the existing DCPP RTS/ESFAS setpoints, the

existing difference between the safety analysis limit and the

setpoints was sufficient to accommodate any changes in instrument

uncertainty.

The change in the pressurizer water level - high setpoint does

not affect a safety analysis limit and, therefore, has no effect on

a margin to safety. Since the normal pressurizer level is maintained

at 60 percent span and the no-load Tavg control level is 22

percent span, a change in the setpoint from less than or equal to 92

percent span to less than or equal to 90 percent span is not

significant to either DCPP plant operation or safety.

The change in the RCS loss of flow-low safety analysis limit

from 87 percent MMF to 85 percent MMF does not affect the existing

plant setpoint and was evaluated to have a negligible effect on the

limiting conditions of a partial loss of flow accident, a single RCP

locked rotor, or RCP shaft break accident. This safety limit change

was also found to have no effect on the DCPP minimum DNBR since the

minimum DNBR is associated with the complete loss of flow accident.

The complete loss of flow accident was evaluated to the Condition II

fault criteria applicable to the partial loss of flow accident

evaluation and was acceptable.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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

NRC Project Director: William H. Bateman

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: December 23, 1996

Description of amendment request: The proposed amendment would

allow the use of Vantage Plus fuel.

Basis for proposed no significant hazards 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?

The probability of occurrence or the consequences of an accident

previously evaluated is not significantly increased. The VANTAGE +

fuel assemblies containing ZIRLOTM clad fuel rods, thimble and

instrument tubes, IFMs, [intermediate fuel mixing assemblies] and

LPD [low-pressure-drop] mid-grids meet the same fuel assembly and

fuel rod design bases as VANTAGE 5 (without IFMs) fuel assemblies in

the other fuel regions. In addition, the 10 CFR 50.46 criteria will

be applied to the ZIRLOTM clad fuel rods, thimble and

instrument tubes, IFM grids, and LPD mid-grids. The use of these

fuel assemblies will not result in a change to the proposed Indian

Point Unit 3 VANTAGE 5 (without IFMs) transition core design and

safety analysis limits. The ZIRLOTM clad material is similar in

chemical composition and has similar physical and mechanical

properties as that of Zircaloy-4. Thus the cladding integrity is

maintained and the structural integrity of the fuel assembly is not

affected. The ZIRLOTM clad fuel rod improves corrosion

resistance and dimensional stability. In addition, the incorporation

of LPD mid-grids and IFMs improves dimensional stability. Since the

dose predictions in the safety analyses are not sensitive to the

fuel assemblies material changes as specified in this report, the

radiological consequences of accidents previously evaluated in the

safety analyses remain valid. Therefore, the probability or

consequences of an accident previously evaluated is not

significantly increased.

(2) Does the proposed license amendment create the possibility

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

evaluated?

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

accident previously evaluated is not created, since the VANTAGE +

fuel assemblies containing ZIRLOTM clad fuel rods, thimble, and

instrument tubes, IFMs, and LPD mid-grids will satisfy the same

design bases as that used for VANTAGE 5 (w/o IFMs) fuel assemblies

in the other fuel regions. Since the original design criteria is

being met, the ZIRLOTM clad fuel rods, thimble and instrument

tubes, IFMs, and LPD mid-grids will not be an initiator for any new

accident. All design and performance criteria will continue to be

met and no new single failure mechanisms have been created. In

addition, the use of these fuel assemblies does not involve any

alterations to plant equipment or procedures which would introduce

any new or unique operational modes or accident precursors.

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

from any accident previously evaluated is not created.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

The margin of safety is not significantly reduced, since the

VANTAGE + fuel assemblies containing ZIRLOTM clad fuel rods,

thimble and instrument tubes; IFMs, and LPD mid-grids do not change

the proposed Indian Point 3 VANTAGE 5 (w/o IFMs) transition core

design and safety analysis limits. The use of these fuel assemblies

containing fuel rods, thimble and instrument tubes with ZIRLOTM

cladding alloy; IFMs and LPD mid-grids will take into consideration

the normal core operating conditions allowed in the Technical

Specifications. For the transition core and each future cycle reload

core, these fuel assemblies will be specifically evaluated using

standard reload design methods and approved fuel rod design models

and methods. This will include consideration of the core physics

analysis, peaking factors and core average linear heat rate effects.

In addition, the 10 CFR 50.46 criteria will be applied each cycle to

the ZIRLOTM clad fuel rods, thimble and instrument tubes, IFMs,

and LPD mid-grids. Analyses or evaluations will be performed each

cycle to confirm the 10 CFR 50.46 will be met. Therefore, the margin

of safety as defined in the Bases to the Indian Point Unit 3

Technical Specifications and VANTAGE 5 (w/o IFMs) ZIRLOTM

licensing amendment approval is not significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

[[Page 6579]]

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: July 9, 1996 (TXX-96393)

Brief description of amendments: The proposed changes would

increase the minimum allowable value of the Unit 1 Steam Line Pressure-

-Low Safety Injection and Steam Line Isolation functions. These changes

are needed to ensure that the instrumentation error is properly

accounted for in the Technical Specifications.

Basis for proposed no significant hazards 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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The changes in the License Amendment Request proposes more

restrictive setpoint Allowable Values for the Steam Line pressure--

Low channels of the Engineered Safety Features Actuation System

(ESFAS). These more restrictive values assure that all applicable

safety analysis limits are being met. Changing an Allowable Value in

the Technical Specifications has no impact on the probability of

occurrence of any accident previously evaluated. None of the

accident analyses were affected, therefore, the consequences of all

previously evaluated accidents remain unchanged.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed changes involve the use of a more conservative

value for the Allowable Value for the Steam Line Pressure--Low

Safety Injection and Steam Line Isolation functions. As such, none

of the changes effect plant hardware or the operation of plant

systems in a way that could initiate an accident. Therefore, the

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

kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

There were no changes made to any of the accident analyses or

safety analysis limits as a result of this proposed change. Further,

the proposed change does not affect the acceptance criteria for any

analyzed event. ESFAS will remain capable of performing its safety

function, and the new requirement will continue to provide adequate

assurance of that capability. Making the Allowable Value more

restrictive provides increased assurance that the channels will

function within the safety analysis limits assumed in the safety

analyses. The margin of safety established by the Limiting

Conditions for Operation also remains unchanged. Thus there is no

effect on 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: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Project Director: William D. Beckner

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: July 10, 1996 (TXX-96405), as

supplemented by letter dated October 1, 1996 (TXX-96475)

Brief description of amendments: The proposed change would take

credit for the addition of train oriented Fan Coil Units for each UPS &

Distribution Room and would provide redundancy to the existing Air

Conditioning Units.

Basis for proposed no significant hazards 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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The UPS HVAC System is a support system for other safety related

equipment, primarily the Uninterruptible Power Supplies and some of

their distribution equipment. The only impact that this system can

have on the probability or consequences of an accident must result

from the failure of the system to provide adequate support to the

supported safety related equipment when that supported safety

related equipment is required to operate.

Allowing same train cooling to satisfy the LCO is considered

equivalent to the existing Technical Specification. The proposed

changes allow the use of the same train UPS Room Fan Coil Units or

the same train UPS A/C Train to support a UPS & Distribution Room.

Surveillance requirements are added or modified to ensure that

the credited support equipment will be available when needed.

Unnecessary starts of the UPS A/C Trains have been eliminated from

the specifications. Overall, this is considered an enhancement that

will increase the reliability of the UPS HVAC Systems. Because both

the existing specification and the proposed revision to the

specification continue to ensure normal support and the availability

of at least one train of equipment in the event of a design basis

accident, with the same or increased reliability, the consequences

of an accident previously evaluated is not affected.

Changing the specification from a ``common'' specification which

impacts both units simultaneously to a specification which applies

to both units separately is basically just an administrative change.

Having ``common'' specifications is an aid to the operator to

provide an alert that both units are affected. With the new LCO,

both units may not be affected because rooms may now be cooled

separately. Because both CPSES Units remain properly covered,

however, this change will not significantly increase the probability

of consequences of an accident.

The revision to the existing ACTION is considered equivalent

except for the change of the Allowed Outage Time (AOT) from seven

days to 30 days. This change is based on the significance of the

heating and cooling function but does represent an increase in AOT

and thus an increase in the probability that the supported functions

could be unavailable. This increase is not considered significant

based on the following several factors:

a)the systems design is based on a conservative assessment of

the worst postulated conditions in the rooms;

b) generally, less than design cooling is required and a short

duration or partial failure may have little or no impact on the

systems ability to perform its function;

c) the multiple backups available (two UPS A/C Trains and only

one UPS Room Fan Coil Unit per each room) increase the potential of

restoring additional cooling if needed;

d) the ability to perform alternate actions if normal cooling is

lost such as circulating air via existing fans or portable fans

thereby extending the time before cooling must be restored; and

e) the extended AOT would allow more time and opportunity to

perform corrective maintenance to ensure high equipment reliability.

The new ACTION for loss of cooling reflects requirements that

already exist in the Technical Specifications. The AOT for this

ACTION statement is 72 hours which is based on the risk from an

event occurring requiring the inoperable UPS A/C Train, and the

remaining UPS Room Fan Coil Units and A/C Train fans providing the

required protection.

The new ACTION for loss of cooling and ventilation reflects a

conservative response to the potential impact of such a condition.

The proposed AOT is one hour. One hour is based on the time lag

available from the operating temperature to the maximum Technical

Specification limit of the UPS & Distribution Rooms. The addition of

a specific ACTION in lieu of relying on Specification 3.0.3,

although essentially equivalent, is consistent with the methodology

of the improved Standard Technical Specifications and alerts the

operator to the significance of the situation.

The changes made to the surveillance ensure that the UPS Room

Fan Coil Units will operate. The UPS Room Fan Coil Units are

connected to the emergency busses and TS 4.8.1.1.2f. demonstrates

the energization of emergency busses with permanently connected

loads. The changes made to the 18

[[Page 6580]]

month surveillances on the UPS A/C trains were changed from the

Safety Injection signal with the Blackout Test signal to ``...

actual or simulated actuation signal''. This is consistent with

NUREG-1431, ``Standard Technical Specifications Westinghouse

Plants''.

The changes to the BASES are descriptive in nature to reflect

the other changes and by themselves have no impact on the

probability or consequences of an accident.

The ability to cope with station blackout and design basis fires

is maintained or enhanced. For station blackout coping, the UPS A/C

fans are considered to remain available while additional cooling is

provided by a single available Fan Coil Unit.

In summary, the proposed changes take advantage of the increased

reliability offered by the revised system design. It also maintains

the level of support provided by the system while at worst, allowing

a slight decrease in availability (in certain situations) which is

not considered significant. As a result, it is concluded that none

of the changes made to the existing Technical Specification involve

a significant increase in the probability or consequences of an

accident previously evaluated.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any previously evaluated?

Revising this specification to take credit for the new UPS Room

Fan Coil Units, to take credit for same train UPS A/C Train support

for a UPS and Distribution Room, to make the specification unit

specific instead of common, to make the surveillances appropriate

for the credited equipment, and to make the action statements

appropriate for the credited equipment and their significance, does

not by itself alter plant hardware. Plant procedures are only

altered to the extent that the revised specification will allow

different configurations of equipment in the UPS HVAC System to be

operated at different times. These changes ensure continued support

of the safety related equipment in the affected areas and do not

affect the equipments failure or failure modes. As a result, these

changes to the Technical Specification do not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

None of the changes being proposed alter the environmental

conditions which are to be maintained in the areas supported by an

OPERABLE UPS HVAC System during normal operations and following an

accident. As a result, the margin of safety for these functions

remains the same. The only potential adverse impact is the system's

postulated availability, as discussed in the response to question 1

above. This reduction in availability is to a great extent mitigated

by the projected increase in system reliability. As noted in the

response to question 1, there is no significant impact on the

accident analyses. Thus, even if system availability issues were

considered an aspect of margin of safety, the proposed changes do

not involve a significant reduction in 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 Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Project Director: William D. Beckner

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

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

Virginia

Date of amendment request: December 17, 1996

Description of amendment request: The proposed changes will allow

one of the two service water loops to be isolated from the component

cooling water heat exchangers (CCHXs) during power operation in order

to refurbish sections of the isolated service water headers. The

proposed temporary changes will be valid for two periods of up to 35

days each for implementation of the service water upgrades associated

with the repair of the sections of the 24-inch service water supply and

return piping to/from the CCHXs.

Basis for proposed no significant hazards 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 North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The piping refurbishment project and the proposed temporary

changes to the SW [service water] and CC [component cooling]

Technical Specifications have been evaluated to assess their impact

on the normal operation of the SW and CC systems and to ensure that

the design basis safety functions of each system are preserved. The

SW system is required to function during all normal and emergency

operating conditions. During normal plant operation, the SW system

provides cooling water to the CCHXs, charging pump coolers,

instrument air compressor coolers, and control room chiller

condensers of both units. Within the first 168 hour Section 3/

4.7.4.1.d TS AS [Action Statement] of isolation of the header which

is to be repaired, temporary 10'' diameter SW lines (one supply and

one return) will be installed to supply the SW to the charging pumps

coolers, instrument air compressors coolers, Unit 2 CR chillers and

spent fuel pool (SFP) coolers to satisfy design basis conditions.

These temporary lines will be routed from the operating part of the

36'' SW headers while the 24'' headers to CCHXs are being repaired.

The temporary lines will be dismantled when the repaired header is

returned to operation (second 168 hour AS). During the two 35-day

periods, one header will operate with its 24-inch piping to/from the

CCHXs temporarily blanked. To avoid operation of the SW pump at

abnormal conditions (low flow) on this ``partially deadlocked''

header, a temporary cross-connect will be installed to by-pass the

CCHXs.

SW system operation with the cross-connect installed was

evaluated for design basis accident (DBA) conditions. The DBA

condition for the SW system is a loss-of-coolant accident on one

unit with simultaneous loss-of-offsite-power to both units. [An] SW

system hydraulic analysis has been performed to verify that adequate

flow is provided to the containment recirculation spray heat

exchangers (RSHXs) with the temporary cross-connect installed and

throttled open, assuming the occurrence of the most limiting single

failure. Therefore, there is no increase in probability or

consequences of the DBA condition.

Utilizing only one SW header to supply flow to the CCHXs has the

potential to affect the reliability of the CC system and all of the

equipment cooled by CC. A review of the equipment affected by this

phase of the SW restoration project was performed to evaluate the

impact on initiating event frequency. Since the SW system and CC

system are support systems used to remove heat, a failure in either

of these systems does not affect the initiating event frequency of

any design basis event. Additionally, an estimate of the impact on

core damage frequency is provided below. The impact on the North

Anna Probabilistic Safety Assessment (PSA) during implementation of

this DCP [design change package] is similar to impact of work

performed under DCP-94-010 since the scope of work of both DCPs is

repair/replacement of different portions of the same 24'' SW headers

to CCHXs. The only difference from a PSA standpoint is that CDF

[core damage frequency] for DCP-94-010 was calculated based on 140

days supply of CCHXs from one SW header while per this DCP it is

only 70 days. Therefore, results of PSA evaluation for DCP-94-010

are conservatively applied to this DCP. The activities to be

performed during the refurbishment project and the various system

alignments required have been evaluated using the Individual Plant

Examination (IPE) Probabilistic Safety Assessment (PSA) model for

North Anna Power Station. This model is used in a manner that is

generally consistent with the Electric Power Research Institute

(EPRI) PSA Applications Guide TR-105396. The effect on the PSA model

is a slight increase in the frequency of reactor trips and an

increase in the probability of RHR [residual heat removal] failure.

The increased frequency of reactor trips is due to the decreased

reliability of the CC system to supply cooling to the RCP [reactor

coolant pump] motor. When only one SW header is available to the CC

heat exchangers

[[Page 6581]]

the frequency of losing this single header is dominated by the

probability of both SW pumps failing. Also considered was the

frequency of pipe rupture anywhere in the single available header.

When the single SW header fails to supply cooling to the CC heat

exchangers, the CC system will heatup causing inadequate cooling for

sustained operation of the RCPs. Tripping these pumps results in a

reactor trip. The second SW header can be expected to supply other

equipment with cooling. This scenario is appropriately modeled as a

reactor trip with main feedwater available initiating event. A

sensitivity analysis shows the increase in CDF to be about 1E-8/

year. The total effect of this DCP includes a failure analysis of

the reactor coolant pump and motor in case of loss of CCW.

The CC system is also included in the PSA model as a support

system for RHR cooling. The RHR system is used to reduce reactor

coolant system temperatures from 350 deg.F (hot shutdown) to

140 deg.F (cold shutdown). The only accident initiator that requires

the unit to be cooled down and placed on RHR cooling are sequences

which are initiated with a steam generator tube rupture. (Note that,

for the North Anna plant design, RHR is separate from the safety

injection system and the low head safety injection pumps.) The

increased probability for the loss of RHR when only one SW header is

available to the CCHXs is estimated using fault tree analysis and is

dominated by the failure of both SW pumps. The probability for the

loss of both SW pumps aligned to the CCHXs is estimated to be 1.5E-

4. The effect of this increase in RHR failure probability was

determined by adding this probability to the top single event in the

RHR function and recalculating the new CDF. The resulting increase

in CDF as a result of RHR system failure following a steam generator

tube rupture is less than 1E-8 per year.

The CC system is further included in the PSA model as part of

the loss of RCP seal cooling as an initiating event and as a loss of

function during other initiating event scenarios. The effect on the

probability for a loss of RCP seal cooling due to losing CC cooling

to the RCP thermal barriers is negligible due to the high

reliability of the charging system to provide seal injection.

The total effect of this DCP on core damage frequency (CDF) was

estimated by a sensitivity analysis combining both the change in the

reactor trip initiating event frequency and the increased failure

probability of RHR. It was evaluated that during implementation of

this DCP, CCHXs will be supplied from one SW header for 70 days (35

x 2=70), therefore, the increase in CDF previously evaluated in DCP-

94-010 based on 140 days is conservative. This DCP does not affect

the containment systems and there would not be any significant

change in off-site dose since the containment heat removal portion

of the SW system is not affected and the increase in CDF is

insignificant. The small increase in CDF calculated for the repair

activities and the procedure developed to provide contingency

actions result in the conclusion that this work does not represent a

significant increase in core damage frequency.

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

from any accident previously evaluated.

The proposed changes to the allowed outage time only provide

operational flexibility needed to perform necessary repairs. During

the project, there will be a significant time period when all the

CCHXs are aligned to one SW loop. The possibility of an interruption

of SW supply to the heat exchangers during a DBA is eliminated by

defeating the closure of the 24-inch SW isolation MOVs [motor-

operated valves] to the CCHXs on [an] SI/CDA [safety injection/

containment depressurization actuation] signal. Both SW headers will

be available for equipment required for safe shutdown of the units

(i.e., RSHXs, charging pumps, and CR/ESGR [control room/emergency

switchgear room] chillers). The SW pipe repair activities and the

installation/removal of the SW cross-connect and temporary piping do

not create the possibility for a malfunction of equipment different

than previously evaluated. Results of the Johnston Pump NPSH [net

positive suction head] test proved to be satisfactory for the

anticipated SW pump flow rates under modes of station operation for

this project, therefore, the possibility for an accident of a

different type than was previously evaluated in the Safety Analysis

Report will not be created. Based on the above, implementation of

the restoration project and approval of the proposed Technical

Specifications changes will not introduce any new accident

initiators nor affect the performance of accident mitigation

systems.

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

The proposed changes to the schedule only provide operational

flexibility to perform the required SW pipe refurbishment. The

Technical Specifications continue to require the SW and CC systems

to remain functional during the period with a single SW supply to

the CCHXs. As stated in item (1) above, the SW system is fully

capable of performing its DBA function during the course of the pipe

refurbishment project with the proposed Technical Specification

changes in place. The effect of this pipe refurbishment project on

CC system reliability was estimated by a sensitivity analysis

combining both the change in the reactor trip initiating event

frequency and the increased failure probability of RHR resulting in

about a 1E-8 per year increase in CDF. Since this project will not

affect the containment systems, there would not be any significant

change in off-site dose, except that resulting directly from the

slight increase in CDF.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

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

Virginia 23219.

NRC Project Director: F. Mark Reinhart (Acting)

Notice of Issuance of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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: September 10, 1996

Brief description of amendments: The amendments extend the

automatic actuation logic channel functional test interval of the

Engineering Safety

[[Page 6582]]

Features Actuation System and the surveillances test interval of the

containment sump isolation valves from monthly to quarterly.

Date of issuance: January 23, 1997

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

30 days.

Amendment Nos.: 218 and 195

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

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

52963) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated January 23, 1997 No significant

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

Date of application for amendments: October 31, 1996

Brief description of amendments: The amendments relocate the

requirements for seismic monitoring instrumentation from Technical

Specification (TS) 3/4.3.7.2, ``Seismic Monitoring Instrumentation'' to

licensee-controlled documents in accordance with Generic Letter 95-10,

``Relocation of Selected Technical Specifications Requirements Related

to Instrumentation.'' The amendments also add a condition to the

operating licenses which approves the relocation of the TS requirements

to the UFSAR.

Date of issuance: January 29, 1997 Effective date: Immediately, to

be implemented within 90 days.

Amendment Nos.: 117, 102

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications and the license.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66703) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 29, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Power Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: September 9, 1996, as

supplemented December 20, 1996

Brief description of amendment: The amendment revises the Minimum

Channels Operable requirement of Item 4.c (Steam Line Isolation,

Containment Pressure Intermediate--High-High) of Technical

Specification (TS) Table 3.3-3 from 3 channels to 2 channels provided

the provisions of Action Statement 14 are followed. This change makes

this Beaver Valley Power Station, Unit No. 1 TS consistent with the

comparable Beaver Valley Power Station, Unit No. 2 TS. The amendment

also revises the minimum charging pump discharge pressure in TS 3/4.5.5

and associated Bases from 2311 psig to 2397 psig. This change ensures

that safety analysis assumptions for safety injection flow are met.

Date of issuance: January 27, 1997

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No: 201

Facility Operating License No. DPR-66. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55032) The supplemental letter provided clarifying information that did

not change the initial proposed no significant hazards consideration

determination or the original notice. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

January 27, 1997. No significant hazards consideration comments

received: No

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

663 Franklin Avenue, Aliquippa, PA 15001

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: August 15, 1996, and as

supplemented by letters dated October 28, November 15, 1996, and

January 7, 1997.

Brief description of amendment: The amendment changes the Clinton

Power Station (CPS) Technical Specifications to incorporate the revised

Safety Limit Minimum Critical Power Ratio (SLMCPR) as calculated by

General Electric (GE) for CPS Cycle 7. The need to change the SLMCPR

resulted from the 10 CFR Part 21 condition reported by GE in their

letter to the NRC dated May 24, 1996.

Date of issuance: January 22, 1997

Effective date: January 22, 1997

Amendment No.: 113

Facility Operating License No. NPF-62: The amendment revised the

Technical Specifications.

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

FR 47978). The licensee's letters of October 28, November 15, 1996, and

January 7, 1997, provided clarifying information and did not make

significant changes to the initial Federal Register notice. The

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

Safety Evaluation dated January 22, 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

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New London County, Connecticut

Date of application for amendment: August 29, 1996

Brief description of amendment: The amendment revises the Technical

Specifications to (1) modify the applicability requirements for certain

radiation monitors so that the radiation monitors are required to be

operable only when secondary containment integrity is required to be

operable; (2) delineate when secondary containment integrity is

required; (3) modify standby gas treatment operability requirements;

(4) make editorial corrections to clarify the configuration of the

radiation monitors; and (5) revise the associated Bases sections.

Date of issuance: January 14, 1997

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

within 30 days.

Amendment No.: 98

Facility Operating License No. DPR-21: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 17, 1996 (61 FR

54242) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 14, 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, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385

[[Page 6583]]

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

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

Pennsylvania

Date of application for amendments: June 28, 1996, as supplemented

by letters dated November 4 and 5, and December 9, 1996

Brief description of amendments: These amendments revise the

technical specifications to incorporate performance-based testing, in

accordance with 10 CFR Part 50, Appendix J, ``Primary Reactor

Containment Leakage Testing For Water-Cooled Power Reactors,'' Option

B. This option allows utilities to extend the frequencies of the Type A

Containment Leak Rate Test, and Type B and C Local Leak Rate Tests

based on the performance and design of the containment and components.

Date of issuance: January 24, 1997

Effective date: Both units, as of date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 118 and 81

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55038) The supplemental letters provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination or the original notice. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated Janaury 24, 1997. No significant hazards consideration

comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

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

Pennsylvania

Date of application for amendments: May 20, 1996

Brief description of amendments: These amendments revise Technical

Specifications (TS) Sections 3/4.4.9.2, 3/4.9.11.1, 3/4.9.11.2, and the

associated TS Bases 3/4.4.9 and 3/4.9.11, to more clearly describe that

the Residual Heat Removal (RHR) system Shutdown Cooling mode of

operation consists of four ``subsystems.'' These TS sections pertain to

plant operations during Operational Conditions (OPCONs) 4, ``Cold

Shutdown'' and 5, ``Refueling.'' In addition, the proposed TS change

would make administrative changes to TS Section 3/4.4.9.1 to ensure

consistency in terminology regarding the description of Shutdown

Cooling ``subsystems.'' The proposed TS changes are consistent with the

guidance delineated in the Improved TS (i.e., NUREG-1433, Revision 1,

``Standard Technical Specifications General Electric Plants, BWR/4,''

dated April 1995) which indicates that the RHR Shutdown Cooling mode of

operation is comprised of two loops and four subsystems (i.e., two

subsystems per loop).

Date of issuance: January 28, 1997

Effective date: As of date of issuance and shall be implemented

within 30 days.

Amendment Nos.: 119 and 82

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55036) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 28, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: August 5, 1996, as supplemented

December 4, 1996

Brief description of amendment: The amendment revised TS Section

2.1 and its associated TS Bases to reflect the change in the Minimum

Critical Power Ratio Safety Limit due to the plant specific evaluation

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

Station, Unit 2, Cycle 4.

Date of issuance: January 29, 1997

Effective date: As of date of issuance and shall be implemented

within 30 days

Amendment No.: 83

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

Technical Specifications.

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

57491) The December 4, 1996, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination or the initial notice. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated January 29, 1997. No significant hazards consideration comments

received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

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

Pennsylvania

Date of application for amendments: March 29, 1996, as supplemented

by letters dated December 5, 1996, and January 15, 1997

Brief description of amendments: These amendments modify Technical

Specification (TS) Section 4.5.1.d.2.b to delete the requirement to

perform in-situ functional testing of the Automatic Depressurization

System (ADS) valves once every 24-months as part of start-up testing

activities.

Date of issuance: January 29, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 120 and 84

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

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

57488) The December 5, 1996, and January 15, 1997, letters provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination nor the initial notice.

The Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated January 29, 1997. No significant hazards

consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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 application for amendment: October 1, 1996

Brief description of amendment: The amendment allows for a one-time

extension of the surveillance intervals for the containment isolation

valve seat leakage test, the isolation valve seal water test, the boron

injection tank leakage test, the containment spray nozzle test, and the

city water backup to the auxiliary boiler feed pump test.

Date of issuance: January 28, 1997

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

30 days.

Amendment No.: 172

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

[[Page 6584]]

64393) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 28, 1997 No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments: October 1, 1996, supplemented

October 31, 1996

Brief description of amendments: The amendments change Technical

Specifications 3/4.7.1.5, ``Main Steam Line Isolation Valves (MSIVs),''

and 3/4.3.2, ``Engineered Safety Feature Actuation System

Instrumentation.'' The amendments accommodate entry into Modes 3 and 2

prior to performing MSIV closure time testing in Mode 2, allow

additional time for the repair and testing of inoperable MSIVs in

certain operating Modes, delete footnotes that are no longer

applicable, and change the low steam line pressure trip setpoint value

for safety injection, turbine trip and feedwater isolation to make it

consistent with the actual plant configuration.

Date of issuance: January 17, 1997

Effective date: Both units, as of date of issuance, to be

implemented prior to entry into Mode 3 from the current outage.

Amendment Nos. 187 and 170

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55040) The supplemental letter changed the TSs to provide greater

consistency with requirements of NUREG-1431 ``Standard Technical

Specifications - Westinghouse Plants,'' Revision 1, and did not change

the initial proposed no significant hazards consideration determination

or the Federal Register notice. The Commission's related evaluation of

the amendments is contained in a Safety Evaluation dated January 17,

1997. No significant hazards consideration comments received: No

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments: October 24, 1996, as

supplemented December 23, 1996

Brief description of amendments: The amendments changed Technical

Specification 3/4.7.1.2, ``Auxiliary Feedwater System.'' The changes

revised the 18-month surveillance performed on the system's pumps and

valves because testing of the turbine driven Auxiliary Feedwater pump

can only be performed in higher modes when there is sufficient

secondary steam pressure.

Date of issuance: January 23, 1997

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

days

Amendment Nos. 188 and 171

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 19, 1996 (61

FR 58905) The December 23, 1996, letter proposed changes to TS 3/4.3.2

to provide consistency with those proposed in the October 24, 1996,

letter and therefore did not change the initial proposed no significant

hazards consideration determination and was within the scope of the

initial notice. The Commission's related evaluation of the amendments

is contained in a Safety Evaluation dated January 23, 1997. No

significant hazards consideration comments received: No

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments: September 25, 1996

Brief description of amendments: The amendments relocate the list

of containment isolation valves from the Technical Specifications to

the Salem Updated Final Safety Analysis Report and correct references.

Date of issuance: January 30, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 60 days.

Amendment Nos. 189 and 172

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications and the License.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55039) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 30, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

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

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: January 31, 1996, as revised

November 26, 1996. The November 26, 1996, submittal withdrew the

proposed change to surveillance tests being performed at power.

Brief description of amendments: These amendments will revise the

minimum emergency diesel generator day tank fuel oil volume.

Date of issuance: January 17, 1997

Effective date: January 17, 1997

Amendment Nos.: 203 and 184

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7559) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 17, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: : The Alderman Library,

Special Collections Department, University of Virginia,

Charlottesville, Virginia 22903-2498.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of application for amendments: February 8, 1996, as

supplemented August 15, December 2 and December 19, 1996, and January

6, 1997

Brief description of amendments: These amendments revise Technical

Specification (TS) Section 15.3.10, ``Control Rod and Power

Distribution Limits,'' to improve the clarity of this section and add

surveillance requirements to Section 15.4.1, ``Operational Safety

Review.''

Date of issuance: January 16, 1997

Effective date: January 16, 1997, with full implementation within

45 days

Amendment Nos.: Unit 1 - 171, Unit 2 - 175

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: March 13, 1996 (61 FR

10398) The August 15, December 2 and

[[Page 6585]]

December 19, 1996, and January 6, 1997, letters provided clarifying

information and updated TS pages that were within the scope of the

original application and did not change the NRC staff's initial

proposed no significant hazards consideration determination. The

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

Safety Evaluation dated January 16, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point, Units 3 and 4, Dade County, Florida

Date of amendment request: December 17, 1996

Description of amendment request: The proposed amendments would

modify the Turkey Point Units 3 and 4 Technical Specifications to

change the Reactor Coolant Pump (RCP) flywheel surveillance

requirement. The proposed change will require RCP flywheel inspections

once every ten years.

Date of publication of individual notice in Federal Register:

January 10, 1997 (62 FR 1476)

Expiration date of individual notice: February 10, 1997

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,

Illinois

Date of amendment request: December 13, 1996

Description of amendment request: The proposed amendment would

approve transfer of Soyland Power Cooperative's 13.21% minority

ownership interest in the Clinton Power Station to Illinois Power

Company. This action would result in Illinois Power Company becoming

the sole owner of the Clinton Power Station.

Date of publication of individual notice in Federal Register:

January 29, 1997 (62 FR 4337).

Expiration date of individual notice: February 28, 1997

Local Public Document Room location: : Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

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 for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective 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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental

[[Page 6586]]

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 room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By March 14, 1997, the

licensee may file a request for a hearing with respect to issuance of

the amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

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

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

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

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

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

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

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

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. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 20555-001, 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 20555-001, 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 the

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

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

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

Date of application for amendments: January 13, 1997, as

resubmitted January 17, 1997, and supplemented January 22, 1997.

Brief description of amendments: The proposed amendments would:

evaluate the Unreviewed Safety Question (USQ) associated with the

operation of Dresden, Units 2 and 3, with the recently discovered error

in the head loss across the Emergency Core Cooling System (ECCS)

suction strainers; change the Technical Specification (TS) values by

lowering the allowable water temperature in the suppression chamber and

ultimate heat sink; change the basis of the TS to allow credit for two

psig of containment pressure to compensate for a slight increase in the

amount of Net Positive Suction Head (NPSH) deficiency during the first

10 minutes following a design basis accident (DBA); and add a license

condition to allow the licensee to change the Updated Final Safety

Analysis Report to reflect the use of two psig of containment pressure

to compensate for the deficiency in NPSH during the first 10 minutes

following a DBA.

Date of Issuance: January 28, 1997 Effective date: Immediately, to

be implemented within 30 days.

Amendment Nos.: 152/147

Facility Operating License Nos. DPR-19 and DPR-25. The amendments

revised the Technical Specifications and the Operating Licenses. Press

release issued requesting comments as to proposed no significant

hazards

[[Page 6587]]

consideration: Yes January 25, 1997 Joliet Herald News Comments

received: No. The Commission's related evaluation of the amendment,

finding of exigent circumstances, consultation with the State of

Illinois and determination of no significant hazards consideration are

contained in a Safety Evaluation dated January 28, 1997.

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

Austin, One First National Plaza, Chicago, Illinois 60690

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

NRC Project Director: Robert A. Capra

Dated at Rockville, Maryland, this 5th day of February, 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

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

Reactor Regulation

[Doc. 97-3324 Filed 2-11-97; 8:45 am]

BILLING CODE 7590-01-F