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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from July 6, 1996, through July 19, 1996. The

last biweekly notice was published on July 17, 1996 (61 FR 37295).

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 August 30, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

for Domestic Licensing Proceedings'' in 10

[[Page 40014]]

CFR Part 2. Interested persons should consult a current copy of 10 CFR

2.714 which is available at the Commission's Public Document Room, the

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

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

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

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

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

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

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

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

Michigan

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

Description of amendment request: The proposed amendment would

modify the charcoal testing standards for the Control Room Emergency

Filtration System (CREFS) and the Standby Gas Treatment System (SGTS)

to the current industry standard. The changes affect Surveillance

Requirements (SRs) 4.6.5.3.b.2, 4.6.5.3.c, 4.7.2.1.c.2, and 4.7.2.1.d

in Technical Specifications (TS) 3/4.6.5.3 ``Standby Gas Treatment

System'' and TS 3/4.7.2 ``Control Room Emergency Filtration System.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated. By providing an improved protocol for charcoal testing

the proposal provides greater assurance that the installed charcoal

can perform its design function and, thus, the consequences of

evaluated accidents remain valid. The method of laboratory analysis

has no effect upon how the plant is operated, including the method

of sample removal. Therefore, the probability [or consequences] of

any evaluated accident is unchanged.

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

new or different kind of accident from any accident previously

evaluated. The proposal has no effect on the manner of plant

operation. The proposal does not involve any change to the plant

design. Therefore, the change creates no new accident modes.

[[Page 40015]]

3. The proposed TS changes do not involve a significant

reduction in a margin of safety. By providing an improved protocol

for charcoal testing the proposal acts to maintain existing safety

margins. The change to the SGTS charcoal acceptance criteria also

acts to ensure that the existing margins, as discussed in Regulatory

Guide 1.52, Revision 2 [Design, Testing and Maintenance Criteria for

Post-Accident Engineered Safety-Feature Atmosphere Cleanup System

Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear

Power Plants], are maintained.

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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226

NRC Project Director: Mark Reinhart, Acting

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: June 18, 1996

Description of amendment request: For Beaver Valley Power Station,

Unit No. 1 (BVPS-1) only, the proposed amendment would revise Technical

Specification (TS) 3.4.5 and associated Bases; the Bases for TS 3.4.6.2

would also be revised. The proposed changes are editorial in nature and

are intended to provide consistency between the TSs and associated

Bases. Index page XIX would be revised to reflect the revision of page

numbers for TS Tables 4.4-1 and 4.4-2 due to shifting of text.

For Beaver Valley Power Station, Unit No. 2 (BVPS-2) only, the

proposed amendment would implement a voltage-based repair criteria for

steam generator tubes similar to the changes approved for BVPS-1 by

License Amendment No. 198. The proposed changes are intended to reflect

the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based

Repair Criteria for Westinghouse Steam Generator Tubes Affected by

Outside Diameter Stress Corrosion Cracking.'' The proposed changes

would revise TSs 3.4.5 and 3.4.6.2 and associated Bases. TS Table 4.4-2

would be revised to reference TS 6.6 for reporting requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Tube burst criteria are inherently satisfied during normal

operating conditions due to the proximity of the tube support plate

(TSP). Test data indicates that tube burst cannot occur within the

TSP, even for tubes which have 100% throughwall electric discharge

machining notches, 0.75 inch long, provided that the TSP is adjacent

to the notched area. Since tube-to-TSP proximity precludes tube

burst during normal operating conditions, use of the criteria must

retain tube integrity characteristics which maintain a margin of

safety of 1.43 times the bounding faulted condition, main steamline

break (MSLB) pressure differential. The Regulatory Guide (RG) 1.121

criterion requiring maintenance of a safety factor of 1.43 times the

MSLB pressure differential on tube burst is satisfied by 7/8''

diameter tubing with bobbin coil indications with signal amplitudes

less than 8.6 volts, regardless of the indicated depth measurement.

The upper voltage repair limit (VURL) will be determined

prior to each outage using the most recently approved NRC database

to determine the tube structural limit (VSL). The structural

limit is reduced by allowances for nondestructive examination (NDE)

uncertainty (VNDE) and growth (VGR) to establish

VURL. Using the Generic Letter (GL) 95-05 NDE and growth

allowances for an example, the NDE uncertainty component of 20% and

a voltage growth allowance of 30% per full power year can be

utilized to establish a VURL of 5.7 volts. The 20% NDE

uncertainty represents a square-root-sum-of-the-squares (SRSS)

combination of probe wear uncertainty and analyst variability. The

degradation growth allowance should be an average growth rate or 30%

per effective full power year, whichever is larger.

Relative to the expected leakage during accident condition

loadings, it has been previously established that a postulated MSLB

outside of containment but upstream of the main steam isolation

valve (MSIV) represents the most limiting radiological condition

relative to the plugging criteria. In support of implementation of

the revised plugging limit, analyses will be performed to determine

whether the distribution of cracking indications at the tube support

plate intersections during future cycles are projected to be such

that primary-to-secondary leakage would result in postulated site

boundary and control room doses exceeding 10 CFR 100, 10 CFR 50

Appendix A, and GDC-19 [General Design Criterion-19] requirements,

respectively. A separate calculation has determined the maximum

allowable MSLB leakage limit in a faulted loop. This limit was

calculated using the technical specification reactor coolant system

(RCS) Iodine-131 activity level of 1.0 microcuries per gram dose

equivalent Iodine-131 and the recommended Iodine-131 transient

spiking values consistent with NUREG-0800. The projected MSLB

leakage rate calculation methodology prescribed in Section 2.b of GL

95-05 will be used to calculate the end-of-cycle (EOC) leakage.

Projected EOC voltage distribution will be developed using the most

recent EOC eddy current results and considering an appropriate

voltage measurement uncertainty. The log-logistic probability of

leakage correlation will be used to establish the MSLB leakrate used

for comparison with the faulted loop allowable limit. Therefore, as

implementation of the voltage-based repair criteria does not

adversely affect steam generator tube integrity and implementation

will be shown to result in acceptable dose consequences, the

proposed amendment does not result in any increase in the

probability or consequences of an accident previously evaluated in

the Updated Final Safety Analysis Report (UFSAR).

The proposed changes to the BVPS-1 Index, Specifications and

associated Bases and the proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. Therefore, these changes do not involve an

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

Implementation of the proposed steam generator tube voltage-

based repair criteria does not introduce any significant changes to

the plant design basis. Use of the voltage-based repair criteria

does not provide a mechanism which could result in an accident

outside of the region of the tube support plate elevations as no

outside diameter stress corrosion cracking (ODSCC) is occurring

outside the thickness of the tube support plates. Neither a single

or multiple tube rupture event would be expected in a steam

generator in which the plugging limit has been applied (during all

plant conditions).

Duquesne Light Company will implement a maximum primary-to-

secondary leakage rate limit of 150 gpd [gallons per day] per steam

generator to help preclude the potential for excessive leakage

during all plant conditions. The RG 1.121 criterion for establishing

operational leakage rate limits that require plant shutdown are

based upon leak-before-break considerations to detect a free span

crack before potential tube rupture during faulted plant conditions.

The 150 gpd limit provides for leakage detection and plant shutdown

in the event of the occurrence of an unexpected single crack

resulting in leakage that is associated with the longest permissible

crack length. RG 1.121 acceptance criteria for establishing

operating leakage limits are based on leak-before-break

considerations such that plant shutdown is initiated if the leakage

associated with the longest permissible crack is exceeded.

The single through-wall crack lengths that result in tube burst

at 1.43 times the MSLB pressure differential and the MSLB pressure

differential alone are approximately 0.57 inch and approximately

0.84 inch, respectively. A leak rate of 150 gpd will provide for

detection of approximately 0.41 inch long cracks at nominal leak

rates and approximately 0.62 inch long cracks at the lower 95%

confidence level leak rates. Since tube burst is precluded during

normal

[[Page 40016]]

operation due to the proximity of the TSP to the tube and the

potential exists for the crevice to become uncovered during MSLB

conditions, the leakage from the maximum permissible crack must

preclude tube burst at MSLB conditions. Thus, the 150 gpd limit

provides for plant shutdown prior to reaching critical crack lengths

for MSLB conditions using the lower 95% leakrate data. Additionally,

this leak-before-break evaluation assumes that the entire crevice

area is uncovered during blowdown. Partial uncovery will provide

benefit to the burst capacity of the intersection. Analyses have

shown that only a small percentage of the TSPs are deflected greater

than the TSP thickness during a postulated MSLB.

As steam generator tube integrity upon implementation of the

voltage-based repair criteria continues to be maintained through

inservice inspection and primary-to-secondary leakage monitoring,

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

accident previously evaluated is not created.

The proposed change to BVPS-1 Index, Specifications and

associated Bases and the proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. These changes do not change the performance of

plant systems, plant configuration or method of operating the plant.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The use of the voltage-based repair criteria at BVPS-2 maintains

steam generator tube integrity commensurate with the criteria of RG

1.121. This guide describes a method acceptable to the Commission

for meeting GDCs 14, 15, 30, 31, and 32 by reducing the probability

or the consequences of steam generator tube rupture. This is

accomplished by determining the limiting conditions of degradation

of steam generator tubing, as established by inservice inspection,

for which tubes with unacceptable cracking should be repaired or

removed from service. Upon implementation of the proposed criteria,

even under the worst case conditions, the occurrence of ODSCC at the

tube support plate elevations is not expected to lead to a steam

generator tube rupture event during normal or faulted plant

conditions. The EOC distribution of crack indications at the tube

support plate elevations will be confirmed to result in acceptable

primary-to-secondary leakage during all plant conditions and that

radiological consequences remain within the licensing basis.

In addressing the combined effects of loss-of-coolant-accident

(LOCA) + safe shutdown earthquake (SSE) on the steam generator

component (as required by GDC 2), it has been determined that tube

collapse may occur in the steam generators at some plants. This is

the case as the tube support plates may become deformed as a result

of lateral loads at the wedge supports at the periphery of the plate

due to the combined effects of the LOCA rarefaction wave and SSE

loadings. Then, the resulting pressure differential on the deformed

tubes may cause some of the tubes to collapse. There are two issues

associated with steam generator tube collapse. First, the collapse

of steam generator tubing reduces the 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 peak clad temperature. Second, there is a potential that

partial through-wall cracks in tubes could progress to complete

through-wall cracks during tube deformation or collapse.

The results of an analysis using the larger break inputs show

that the LOCA loads were found to be of insufficient magnitude to

result in steam generator tube collapse or significant deformation.

Since the leak-before-break methodology is applicable to the reactor

coolant loop piping, the probability of breaks in the primary loop

piping is sufficiently low that they need not be considered in the

structural design of the plant. The limiting LOCA event becomes the

pressurizer spray line break. Analysis results have demonstrated

that no tubes were subject to deformation or collapse. No tubes have

been excluded from application of the subject voltage-based steam

generator tube repair criteria.

Addressing RG 1.83 considerations, implementation of the

voltage-based repair criteria is supplemented by: enhanced eddy

current inspection guidelines to provide consistency in voltage

normalization, the bobbin coil inspection will include 100% of the

hot-leg TSP intersections and cold-leg intersections down to the

lowest cold-leg TSP with known ODSCC, the determination of the TSPs

having ODSCC will be based on the performance of at least 20% random

sampling of tubes inspected over their full length, and rotating

pancake coil inspection requirements for the larger indications left

inservice to characterize the principal degradation as ODSCC.

As noted previously, implementation of the tube support plate

intersection voltage-based repair criteria will decrease the number

of tubes which must be repaired. The installation of steam generator

tube plugs reduces the RCS flow margin. Thus, implementation of the

voltage-based repair criteria will maintain the margin of flow that

would otherwise be reduced in the event of increased tube plugging.

The proposed change to the BVPS-1 Index, Specifications and

associated Bases and the proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. These changes will not reduce the margin of

safety because they have no impact on any safety analysis

assumptions.

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

amendment request does not result in a significant reduction in

margin with respect to plant safety as defined in the UFSAR or any

BASES of the plant technical specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001

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

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

NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: May 9, 1996

Description of amendment request: The proposed amendment changes

both technical and administrative requirements associated with station

batteries. The proposed changes are modeled after ``Standard Technical

Specifications - Babcock and Wilcox Plants,'' NUREG-1430 and Nuclear

Energy Institute guidance, ``IEEE Recommended Practice for Maintenance,

Testing, and Replacement of Vented Lead-Acid Batteries for Stationary

Applications,'' IEEE Std 450-1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The switchyard 125V DC control power source requirements do not

meet the criteria for inclusion in Technical Specifications (TSs) as

evaluated with respect to the selection criteria of 10 CFR 50-36.

These control power sources are not assumed to mitigate accident or

transient events. The effects of a loss of these control power

sources are enveloped by the Loss of Offsite Power (LOOP) event and

relocation is considered to have a non-significant impact on the

probability or severity of a LOOP event. These requirements will be

relocated from the TSs to an appropriate administratively controlled

document and maintained pursuant to 10 CFR 50.59.

Proposed changes incorporating the requirements of TS 3.7.1.D,

3.7.2.E, 3.7.2.F, and 3.7.2.A, as related to the DC electrical power

subsystems in the new TS 3.7.3 results in a more stringent

requirement for the ANO-1 TSs in that reductions to lower conditions

of operation in shorter periods of time are now required. These more

stringent requirements are not assumed to be initiators of any

analyzed events and will not alter assumptions relative to

mitigation of accident or transient events.

Proposed changes incorporating TS 3.7.4. requirements for the

station batteries allowing the battery parameters to be outside

[[Page 40017]]

the limits of the Battery Inspection Program for 31 days do not

result in an increase in the frequency of consequences of any

analyzed accident, as the actions require more frequent checks of

other parameters to ensure battery capability during this 31 day

period. The Battery Inspection Program also requires evaluations to

determine battery operability in the event these limits are

exceeded. If an evaluation shows the battery is incapable of

performing its design basis function, that DC electrical subsystem

will be declared inoperable, and the appropriate actions taken.

Proposed changes to allow the use of float current in lieu of

specific gravity incorporate current industry guidance on

operability measures for station batteries, as stated in IEEE-450,

``IEEE Recommended Practice for Maintenance, Testing, and

Replacement of Vented Lead-Acid Batteries for Stationary

Applications.'' This Surveillance Requirement is not considered to

initiate or mitigate any analyzed accident.

The proposed incorporation of a Battery Inspection Program

relocates maintenance requirements from the TSs to a program under

10 CFR 50.59 control and allows the TSs to concentrate on those

items required to ensure battery operability. These relocated

requirements are not considered to be initiators of any analyzed

accident. Battery operability is assured by the combination of TS

Surveillance Requirements and Battery Inspection Program maintenance

requirements based on IEEE-450 guidance.

Proposed changes in Surveillance Requirements and Frequencies

reflect current industry guidance on maintenance and testing of the

station batteries. These requirements, in themselves, are not

considered to be initiators of any analyzed accident condition.

Although some frequencies have been extended, continued performance

of maintenance activities in accordance with IEEE-450, in addition

to the required Surveillance Requirements, ensures that corrective

maintenance can be performed prior to a condition challenging an

operability limit.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The proposed changes do not change the design, configuration, or

method of operation of the plant.

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

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

Relocation of the switchyard 125V DC control power source

requirements has no impact on any safety analysis assumptions. In

addition, the requirements associated with these control power

sources are relocated to an owner controlled document for which

future changes will be evaluated pursuant to the requirements of 10

CFR 50.59.

Proposed changes incorporating the requirements of TS 3.7.1.D,

3.7.2.E, 3.7.2.F, and 3.7.2.A, as related to the DC electrical power

subsystems, in the new TS 3.7.3 impose more stringent requirements

than previously specified for ANO-1.

Proposed changes incorporating TS 3.7.4 requirements for the

station batteries allowing the battery parameters to be outside the

limits of the Battery Inspection Program for 31 days may involve an

incremental reduction in the margin of safety since the battery may

be in a slightly degraded state. However, this reduction is not

considered significant in that the associated actions require more

frequent checks of other parameters to ensure battery capability

during this 31 day period. The attery Inspection Program also

requires evaluations to determine battery operability in the event

these limits are exceeded.

If an evaluation shows the battery is incapable of performing

its design basis function, that DC electrical subsystem will be

declared inoperable, and the appropriate actions taken.

The proposed change to allow the use of float current in lieu of

specific gravity as a measure of battery operability is expected to

result in a more representative measure of operability. IEEE-450

states that specific gravity may not be an appropriate measure of

battery capability following addition of electrolyte or when the

battery is on recharge following a discharge.

Proposed incorporation of a Battery Inspection Program relocates

maintenance requirements from the TSs to a program under 10 CFR

50.59 controls and allows the TSs to concentrate on those items

required to ensure battery operability. The relocation of these

requirements is not considered to be a reduction in the margin of

safety. Battery operability is assured by the combination of TS

Surveillance Requirements and Battery Inspection Program maintenance

requirements based on IEEE-450 guidance.

Proposed changes in Surveillance Requirements and Frequencies

reflect current industry guidance on maintenance and testing of the

station batteries. Although some frequencies have been extended,

continued performance of maintenance activities in accordance with

IEEE-450, in addition to the required Surveillance Requirements,

ensures that corrective maintenance can be performed prior to a

condition challenging an operability limit.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: June 27, 1996

Description of amendment request: The proposed amendment will

modify Technical Specification 3/4.3.3.6, ``Accident Monitoring

Instrumentation,'' based on the Combustion Engineering improved

Standard Technical Specifications (STS) issued by the NRC as NUREG

1432. The amendment will also revise the Technical Specification (TS)

to include Accident Monitoring Instrumentation as recommended by

Regulatory Guide (RG) 1.97, Revision 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change deletes all non-Type A and non-Category 1

instruments from the requirements of TS 3/4.3.3.6, ``Accident

Monitoring Instrumentation.'' Type A variables provide the primary

information required to permit the control room operators to take

specific manually controlled actions, for which no automatic control

is provided, that are required for safety systems to accomplish

their safety functions during a DBA [Design Basis Accident].

Category 1, non-Type A variables are important in reducing public

risk and are retained in TS because they are intended to assist

operators in minimizing the consequences of accidents. Category 2

instruments are generally designated for indicating system operating

status and are not designated as essential key variables necessary

for the safe shutdown of the plant. The proposed change preserves

the safety requirements of RG 1.97, Revision 3, and will not

adversely affect any material condition of the plant that could

directly contribute to causing or mitigating the affects of an

accident.

The proposed change also adds two parameters to TS 3/4.3.3.6

which were previously controlled administratively or per another TS.

Containment Pressure (Wide Wide Range) is being added because it is

a Category 1 parameter required in addition to Containment Pressure

(Wide Range), which is currently in the TS. Neutron Flux is being

added to distinguish the RG 1.97 channels from the non-RG 1.97

channels and to provide action and surveillance requirements

consistent with the other accident monitoring instrumentation. These

additions to TS 3/4.3.3.6 contribute to the overall safety of the

plant and therefore in no way increase the probability or

consequences of an accident previously evaluated.

Additionally, the proposed change also extends the AOTs [Allowed

Outage Times] for TS 3/4.3.3.6 and replaces the HOT SHUTDOWN

requirement for the number of OPERABLE channels being less than the

Required Number of channels with a Special Report requirement. These

changes are based

[[Page 40018]]

on the relatively low probability of an accident occurring which

would require these instruments, the passive nature of these

instruments, and alternate means of monitoring available. This is

consistent with the CE improved STS and associated safety analyses

which have been approved and issued by the NRC as NUREG 1432.

The remainder of the proposed change provides enhancements and

clarifications to TS 3/4.3.3.6 which have no potential to impact

plant operations. No previous accident scenario is changed, and

initiating conditions and assumptions remain as previously analyzed.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed change will not alter the operation of the plant or

the manner in which the plant is operated. No new or different

failure modes have been introduced. TS 3/4.3.3.6 ensures the

OPERABILITY of essential Post Accident Monitoring Instrumentation.

This instrumen-tation provides information to the control room

operators during an accident so that appropriate actions can be

taken to mitigate the consequences of the accident. These

instruments are passive in nature in that no critical automatic

action is assumed to occur from these instruments. Therefore, the

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change revises TS 3/4.3.3.6 based on the

information provided in CE improved STS, NUREG 1432. The deletion

and addition of specific components from the TS per this change is

commensurate with the safety significance of their associated

parameters. The proposed change ensures the operability of the post

accident monitoring instrumentation which has been designated, by RG

1.97 and Waterford 3's associated analysis, as essential for

availability during and following a DBA. The proposed change

preserves the single failure criteria required for this

instrumentation and maintains the level of safety currently

established in the Technical Specifications. The proposed change

will not affect any physical protective boundary. Therefore, the

proposed change will not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: July 17, 1996 (TSCR 242, Rev. 2)

Description of amendment request: The proposed change to the

Technical Specifications would allow the implementation of 10 CFR Part

50, Appendix J, Option B. This application supersedes the previously

submitted application dated February 23, 1996, which was noticed in the

Federal Register on March 27, 1996 (61 FR 13526).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

GPU Nuclear has determined that this TSCR involves no

significant hazards considerations as defined by NRC in 10 CFR

50.92.

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or occurrence or the consequences of an accident of

malfunction of equipment important to safety as previously evaluated

in the Safety Analysis Report.

The proposed change implements Option B of 10 CFR 50, Appendix J

on performance based containment leakage testing. The proposed

change does not involve a change to the plant design or operation.

Therefore, the proposed change does not affect any of the parameters

or conditions that contribute to initiation of any of the analyzed

accidents or malfunctions. The proposed change does not request an

allowable extension of containment testing. Therefore, a

hypothetical leak could remain undetected for a greater period of

time. This slight increase in risk has been determined to be

insignificant as:

Type A Testing

NUREG 1493 [Performance-Based Containment Leak Test Program]

determined that the effect of containment leakage on overall

accident risk is small as risk is dominated by accident sequences

that result in the failure or bypass of the containment. Industry

wide PCILRTs [primary containment integrated leak rate tests] have

demonstrated that only a small fraction of the leaks discovered

during testing exceeded acceptance criteria, and that the leak rate

has been only marginally above the acceptable limit. Only 3% of all

leaks can be detected only by PCILRT, therefore, only 3% of the

theoretical leaks are affected by the extension to the Type A test

interval. Experience at Oyster Creek agrees with the industry wide

data in that the majority of the detected leakage from the primary

containment is found through Type B and C testing. NUREG 1493 found

that these observations, together with the insensitivity of reactor

accident risk to the containment leakage rate, demonstrates that

increasing the Type A leakage test intervals would have a minimal

impact on public risk.

Type B and C Testing

Penetrations are designed to ensure reliability of the

containment isolation function. Type B penetrations use a double

passive seal (e.g. o-ring, gasket) and Type C penetrations use a

double isolation valve design to ensure reliability of the isolation

function. Because valves perform the isolation function actively,

they are more likely to fail on demand (e.g. failure to completely

close on demand). To address this failure mode, Type C valves are

subjected to increased design constraints and testing to ensure both

acceptable leak rates and stroke times. The proposed change does not

alter the installation, operation, operating environment, or testing

method of these valves. Therefore, the proposed change does not

introduce any new component failure modes, nor does it affect the

probability of occurrence of any existing evaluated failure mode.

The failure of any single penetration barrier (isolation valve

or passive seal) does not cause penetration failure. Therefore, a

double failure would have to occur to cause a failure of the

penetration and affect containment. Additionally, the proposed

change does not change the acceptance criteria for acceptable

leakage testing.

The proposed change does not alter plant design or operation,

nor does it alter the allowable maximum leakage rate limit. Thus,

the proposed change does not affect the probability of occurrence

nor the consequences of any evaluated accident or malfunction of

equipment important to safety.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of any accident or

malfunction different from any accident or malfunction previously

evaluated.

The proposed change does not involve a change to the plant

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

affect any of the parameters or conditions that could contribute to

initiation of any accidents. This change only involves the reduction

in Type A, B, and C test frequencies, and the Type A test pressure.

Type A Testing

The only changes proposed to the Type A testing are to frequency

and test pressure. As the proposed test pressure is greater than the

existing test pressure, no new type of accident or malfunction is

created, and the increase in pressure provides an additional margin

of safety. The increase in surveillance interval cannot introduce

any new type of accident or malfunction.

The PCILRT is presently performed at 20 psig. Performance of the

PCILRT at PGG5Ga(35 psig) will provide a more direct leak rate for

analysis. Pa is the design pressure of the torus (the drywell

design pressure is 44 psig, but the torus is non isolable from the

drywell). Therefore, Pa will not create the possibility of the

failure of the torus due to overpressurization. No new accident

modes can be created by extending the test intervals. No safety

related functions

[[Page 40019]]

or components are altered as a result of this change. Therefore, no

new accident or malfunction different from those evaluated in the

Safety Analysis Report can result due to the increase in test

pressure or increase in surveillance interval.

Type B and C Testing

The proposed change only deals with the frequency of performing

Type B and C testing. It does not change what components are tested

or the method of testing. There is no proposed change to the design

or operation of the plant. Therefore, no new accident or malfunction

different from those evaluated in the Safety Analysis Report can

result due to the increase in test pressure or increase in

surveillance interval.

3. Operation of the facility in accordance with the proposed

amendment would not decrease the margin of safety as defined in the

bases of the Technical Specifications.

Type A Testing

Except for the method of defining the test frequency and

pressure at which the PCILRT is performed, the methods for

performing the actual test are not changed. However, the proposed

change can increase the probability that an increase in leakage

could go undetected for an extended period of time. NUREG 1493 has

determined that under several different accident scenarios, the

increased risk of radioactivity release from containment is

negligible with the implementation of these proposed changes.

Type B and C Testing

The proposed change only affects the frequency of Type B and C

testing. The methods for performing the actual test are not changed.

The design or operation of Type B and C components are not changed.

The proposed change will result in a longer interval between tests

of good performing Type B and C components.

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

the proposed change involves the offsite dose consequences of

postulated accidents which are directly related to containment

leakage rate. The containment isolation system is designed to limit

leakage to La, which is defined by the Oyster Creek Technical

Specifications to be 1.0 percent by weight of the containment air at

35 psig per 24 hours. The limitation on containment leakage rate is

designed to ensure the total leakage volume will not exceed the

value assumed in the accident analyses at the peak accident pressure

(Pa). The margin of safety for the offsite dose consequences of

postulated accidents directly related to the containment leakage

rate is maintained by meeting the 1.0 La acceptance criteria.

The La value is not being modified by this proposed Technical

Specification change request.

Therefore, the margin of safety as defined in the bases for the

Technical Specification will not be reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz

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

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

Date of amendment request: June 28, 1996

Description of amendment request: This amendment would allow

implementation of Option B to 10 CFR Part 50, Appendix J, which permits

performance based determination of the frequency of containment leak

rate testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

The proposed change has been evaluated against the standards in

10 CFR 50.92 and determined not to involve a significant hazards

consideration, in that the editorial changes do not change the

meaning or intent of the technical specifications, and operation of

the facility in accordance with the proposed amendment.

1. Would not involve a significant increase in the probability

of occurrence or the consequences of an accident previously

evaluated, because the proposed changes are either purely

administrative changes (involving format, wording, or reporting

requirements) or changes in containment leakage test requirements

(minor scope changes or increased intervals between containment

leakage tests). None of these changes are related to conditions

which cause accidents. The proposed changes do not involve a change

to the plant design or operation.

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

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

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

expected leakage from containment and the associated consequences.

The increased risk due to lengthening of the intervals between

leakage tests was also evaluated and found to be acceptable. Using a

statistical approach, NUREG-1493 determined the increase in the

expected dose to the public from extending the testing frequency to

be extremely small.

2. Would not create the possibility of a new or different kind

of accident from any accident previously evaluated, because the

testing or reporting requirements associated with this change do not

involve a physical alteration of the plant design or changes in the

methods governing normal plant operation. No safety related

equipment or safety related functions are altered as a result of

this change. As a result, the proposed change does not affect any of

the parameters or conditions that could contribute to initiation of

any accidents.

3. Would not involve a significant reduction in a margin of

safety because the proposed changes are either purely administrative

(involving format, wording, or reporting requirements) or changes in

containment leakage test requirements (minor scope changes or

increased intervals between containment leakage tests) such that the

allowable containment leakage rates presently specified in the

Technical Specifications remain unchanged. The Technical

Specifications and the Reactor Building Leakage Rate Testing Program

will ensure that containment system testing is performed in full

compliance with 10 CFR 50 Appendix J.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location:

Law/Government Publications Section, State Library of

Pennsylvania, (REGIONAL DEPOSITORY) Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: John F. Stolz

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

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

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

Matagorda County, Texas

Date of amendment request: May 1, 1995, as supplemented by letters

dated June 22, August 28, November 22, and December 19, 1995, and

January 4, 8 (two letters), and 23, June 27, and July 9, 1996.

Description of amendment request: The proposed amendment would

allow extension of the standby diesel generator allowed outage time to

14 days, and extension of the essential cooling water loop and the

essential chilled water loop allowed outage times to 7 days. The

proposed change would also add to Administrative Controls a description

of the Configuration Risk Management Program (CRMP) used to assess

changes in core damage probability resulting from applicable plant

configurations. This application was previously published in the

Federal

[[Page 40020]]

Register on February 8, 1996, (61 FR 4805).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Standby Diesel Generators are not accident initiators,

therefore the increase in Allowed Outage Times for this system does

not increase the probability of an accident previously evaluated.

The three train design of the South Texas Project ensures that even

during the seven days the Essential Cooling Water loop or the

Essential Chilled Water loop is inoperable there are still two

complete trains available to mitigate the consequences of any

accident. If the Essential Cooling Water and the Essential Chilled

Water loops are operable during the 14 days the Standby Diesel

Generator is inoperable, the Engineered Safety Features bus and

equipment in the train associated with the inoperable Standby Diesel

Generator will be operable. This ensures that all three redundant

safety trains of the South Texas Project design are operable. In

addition the Emergency Transformer will be available to supply the

Engineered Safety Features bus normally supplied by the inoperable

Standby Diesel Generator. These actions will ensure that the changes

do not involve a significant increase in the consequences of

previously evaluated accidents.

The addition of the Configuration Risk Management Program to the

Administrative Section of the Technical Specifications does not

affect current accident analyses.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes affect only the magnitude of the Standby

Diesel Generator, Essential Cooling Water and the Essential Chilled

Water Allowed Outage Times as identified by the marked-up Technical

Specification. As indicated above, the proposed change does not

involve the alteration of any equipment nor does it allow modes of

operation beyond those currently allowed. Therefore, implementation

of these proposed changes does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes result in no significant increase in core

damage or large early release frequencies. Three sets of PSA

[probabilistic safety assessment] results have been presented to the

NRC for the South Texas Project. One submitted in 1989 from the

initial Level 1 PSA of internal and external events with a mean

annual average CDF [core damage frequency] estimate of 1.7E-4, a

second one submitted in 1992 to meet the IPE [individual plant

examination] requirements from the Level 2 PSA/IPE with a CDF

estimate of 4.4E-5, and an update of the PSA that was reported in

the August 1993 Technical Specifications submittal with a variety of

CDF estimates for different assumptions regarding the rolling

maintenance profile and different combinations of modified Technical

Specifications. The South Texas Project PSA was updated in March of

1995 to include the NRC approved Risk-Based AOTs [allowed outage

times] and STIs [surveillance test intervals], Plant Specific Data

and incorporate the Emergency Transformer into the model. This

update resulted in a CDF estimate of 2.07E-5 per reactor year. When

the requested changes are modeled, the resulting CDF estimate is

2.18E 10-5 (sic) [2.18E-5] per reactor year. This corresponds to

5.2% decrease in the Core Damage Frequency calculated for the

previously submitted 21 Day AOT. The Large, Early Release Frequency

is quantified as 4.69E-07 per reactor year which represents a

decrease of 7.5% from the value calculated for the previously

submitted 21 Day AOT. Therefore, it is concluded that there is no

significant reduction in the margin of safety.

Based on the above evaluation, the South Texas Project has

concluded that these changes do not involve a significant hazards

consideration.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: William D. Beckner

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,

Linn County, Iowa

Date of amendment request: July 5, 1996

Description of amendment request: The proposed Technical

Specification (TS) amendment would support implementation of Noble

Metal Chemical Addition (NMCA) at the Duane Arnold Energy Center (DAEC)

as a method to enhance the effectiveness of Hydrogen Water Chemistry

(HWC) in mitigating Intergranular Stress Corrosion Cracking (IGSCC) in

Boiling Water Reactor (BWR) vessel internal components. The proposed

amendment would raise the reactor water conductivity limit in STARTUP

and HOT SHUTDOWN only during the application of NMCA. The reactor water

conductivity will be restored after the NMCA.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed TS amendment will not significantly increase the

probability or consequences of any previously evaluated accidents.

It is expected that during the NMCA application period, the

reactor water conductivity will increase and exceed the conductivity

limit of 2.0 [micro]mhos/cm specified in our current TS. Our current

TS requires that whenever the reactor is in STARTUP or HOT SHUTDOWN

Mode, the conductivity shall not exceed 2.0 [micro]mhos/cm for more

than 48 continuous hours or be in HOT SHUTDOWN within the next 12

hours and in COLD SHUTDOWN within the following 24 hours.

The expected increase in conductivity is due to the presence of

noble metal chemistry in the reactor water and is appropriate during

the [NMCA] application period. The deposited layer of noble metals

is beneficial for mitigating IGSCC in reactor vessel internal

components. Other reactor water chemistry parameters such as

chloride and sulfate are not expected to change; pH is expected to

change but not out of the acceptable range. The reactor water

chemistry parameters will be analyzed to ensure they are within the

normal range, on a frequency consistent with the existing TS,

Sections 4.6.B.2.c and 4.6.B.2.d when conductivity is elevated

during the NMCA application.

During and after the application, the Reactor Water Cleanup

(RWCU) system will continue to operate to remove the excess ions

from the reactor water and restore the reactor water conductivity to

the limit specified in Section 3.6.B. Therefore, this proposed TS

amendment will not significantly increase the probability or

consequences of any previously evaluated accidents.

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

a new or different kind of accident. The proposed TS amendment will

only permit a higher value of the reactor water conductivity limit

during the application period of NMCA. The application is

anticipated to increase the reactor water conductivity.

During and after the application, the RWCU system will continue

to operate to remove the excess ions and restore the reactor water

conductivity to the limit specified in Section 3.6.B. As is

discussed above, the deposited layer of noble metals is beneficial

for mitigating IGSCC in reactor vessel internal components.

Therefore, this proposed TS amendment will not create the

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

previously evaluated.

3. The proposed TS amendment will only permit a higher value of

the reactor water conductivity limit during the application period

of NMCA. The increase in

[[Page 40021]]

conductivity is anticipated during the application and is

appropriate. The deposited layer of noble metals is beneficial for

mitigating IGSCC in reactor vessel internal components. During and

after the application, the RWCU system will continue to operate to

remove the excess ions and restore the reactor water conductivity to

the limit specified in Section 3.6.B. Therefore, no margin of safety

is reduced as a result of the anticipated increase in conductivity

due to the addition of the known noble metals.

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

thisreview, 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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,

Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869

NRC Project Director: Gail H. Marcus

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: June 21, 1996

Description of amendment request: The proposed amendment would

modify Section 5.7, ``High Radiation Areas,'' of the ``Administrative

Controls'' section of the Clinton Power Station technical

specifications (TS). The proposed changes include: (1) allowing

utilization of a Radiation Work Permit (RWP) ``or equivalent'' to

control entry into a high radiation area; (2) clarifying the example

given in the TS of individuals who are qualified in radiation

protection procedures; (3) clarifying the requirements for when

specified access controls and barriers for high radiation areas within

large areas like the containment must be established; (4) clarifying

that it is acceptable for an RWP to specify a maximum dose, i.e., a

specified setpoint on an alarming dosimeter in lieu of a stay time for

entry into a high radiation area (where an individual could receive a

deep dose equivalent of 3000 mrem in one hour); (5) eliminating the

upper dose limit for specifying the applicability of the requirements

of Specification 5.7.1; (6) providing additional flexibility regarding

who may control the keys to locked doors for preventing unauthorized

entry into high radiation areas; (7) reorganizing TS Sections 5.7.1,

5.7.2, and 5.7.3 into four sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4);

and (8) making minor edits to enhance readability.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

(1) None of the proposed changes involve a significant increase

in the probability or consequences of any accident previously

evaluated.

The proposed changes do not change the design or the operation

of the plant. The proposed changes are only related to the control

of access to high radiation areas for the purpose of controlling

dose to plant personnel. Because no change to plant design is

proposed, there is no impact to any accident mitigating system.

Likewise, because there is no proposed change to plant operating

procedures, plant operation is not impacted. This proposed change

does not impact any accident scenario or the previously calculated

post-accident doses. Therefore, the limits of 10 CFR 100 will

continue to be met. No probability or consequence of any accident

previously evaluated is impacted by the proposed changes to TS.

(2) None of the proposed changes create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed amendment is administrative in nature and does not

impact directly or indirectly the design or the operation of the

Clinton Power Station, thus no new accident can be created.

(3) None of the proposed changes involve a significant reduction

in a margin of safety.

There is no reduction to the margin of safety because the

operating limits and functional capabilities of plant safety systems

are unaffected by the proposed changes to administrative

requirements. As noted previously, the proposed changes do not

impact any accident analyses, including the associated dose

calculations. With respect to controls for controlling operational

dose to plant personnel, the proposed changes are intended to

provide clarity and/or flexibility with respect to the

administration and programmatic controls for controlling such dose,

and yet maintain an adequate margin of safety for minimizing dose to

site personnel consistent with the requirements of 10 CFR 20 and

guidance of Regulatory Guide 8.38.

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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th Street,

Decatur, Illinois 62525

NRC Project Director: Gail H. Marcus

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: June 28, 1996

Description of amendment request: The proposed amendment would

allow removal of the Inclined Fuel Transfer System (IFTS) primary

containment blind flange while primary containment is required to be

operable. This will provide flexibility to operate the IFTS for the

purpose of testing and exercising the system during such conditions.

Primary containment integrity will be provided by an alternate means

while the blind flange is removed. The change would be incorporated via

a provisional note into Technical Specification (TS) Surveillance

Requirement 3.6.1.3.3, associated with TS 3.6.1.3, ``Primary

Containment Isolation Valves (PCIVs).''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

(1) The proposed change allows operation of the IFTS while

primary containment operability is required. The proposed change

does not involve any modifications to plant systems or design

parameters or conditions that contribute to the initiation of any

accidents previously evaluated. Therefore, the proposed change

cannot increase the probability of any accident previously

evaluated.

The proposed change potentially affects the leak-tight integrity

of the containment structure which is designed to mitigate the

consequences of a loss-of-coolant accident (LOCA). The function of

the primary containment is to maintain functional integrity during

and following the peak transient pressures and temperatures that

result from any LOCA. The primary containment is designed to limit

fission product leakage following the design basis LOCA. Because the

proposed change does not alter the plant design, only the extent of

the boundaries that provide primary containment isolation for the

IFTS penetration, the proposed change does not result in an increase

in primary containment leakage. However, temporarily using the IFTS

transfer tube and its attached appurtenances as part of the primary

containment boundary (which have not been fabricated or installed to

exactly the same requirements as a fully certified primary

containment penetration) can increase the probability that a LOCA

would challenge the pressure retaining integrity of these

components. Since the subject components have been built to

withstand pressure, temperature, and seismic conditions similar to

those of the existing penetration, they are judged to be an

[[Page 40022]]

acceptable barrier to prevent the uncontrolled release of post-

accident fission products for the purposes of this amendment

request.

Further, it has been shown that the largest potential leakage

pathway, the IFTS transfer tube itself, would remain sealed by the

depth of water required to be maintained in the fuel building fuel

transfer pool. The transfer tube drain line constitutes the other

possible leakage pathway, and will be required to be capable of

being isolated via administrative control of the manual isolation

valve in the drain line. Additionally, due to the physical

relationships of the buildings and components involved, any leakage

from either of these pathways is fully contained within the

boundaries of the secondary containment and would be filtered by the

Standby Gas Treatment System prior to release to the environment.

Based on the above, Illinois Power has concluded that the

proposed change will not result in a significant increase in the

probability or consequences of any accident previously evaluated.

(2) The proposed change does not involve a change to the plant

design or operation (except when the IFTS is operated). As a result,

the proposed change does not affect any of the parameters or

conditions that could contribute to the initiation of any accidents.

No new accident modes are created by this change. Extending the

primary containment boundary to include portions of the IFTS has no

influence on, nor does it contribute to the possibility of a new or

different kind of accident or malfunction from those previously

analyzed.

Based on the above, Illinois Power has concluded that the

proposed change will not create the possibility of a new or

different kind of accident not previously evaluated.

(3) The request does not involve a significant reduction in a

margin of safety. The proposed change only affects the extent of a

portion of the primary containment boundary. Precautions will be

taken to administratively control the IFTS transfer tube drain path

so that the proposed change will not increase the probability that

an increase in leakage from the primary containment to the secondary

containment could occur.

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

the proposed change involves the offsite dose consequences of

postulated accidents which are directly related to containment

leakage rate. The containment isolation system is designed to limit

leakage to La, which is defined by the Clinton Power Station

Technical Specifications to be 0.65% of primary containment air

weight per day at the calculated peak constant pressure (Pa).

The limitation on containment leakage rate is designed to ensure

that total leakage volume will not exceed the value assumed in the

accident analyses at the peak accident pressure (Pa). The

margin of safety for the offsite dose consequences of postulated

accidents directly related to the containment leakage rate is

maintained by meeting the 1.0 La acceptance criteria. The

La value is not being modified by this proposed technical

specification change. The IFTS will continue to provide an

acceptable barrier to prevent containment leakage during a LOCA, and

therefore this change will not create a situation causing the

containment leakage rate acceptance criteria to be violated.

As a result, Illinois Power has concluded that the proposed

change will not result in a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th Street,

Decatur, Illinois 62525

NRC Project Director: Gail H. Marcus

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of amendment requests: June 11, 1996 (AEP:NRC:80027)

Description of amendment requests: The proposed amendments would

remove from the technical specifications (TS) certain requirements for

administrative controls, related to quality assurance requirements, in

accordance with the guidance of NRC Administrative Letter 95-

06,Relocation of Technical Specifications Administrative

Controls Related to Quality Assurance.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

We have evaluated the proposed T/S changes and have determined

that the changes should involve no significant hazards consideration

based on the criteria established in 10 CFR 50.92(c). Operation of

Cook Nuclear Plant in accordance with the proposed amendment will

not satisfy any of the following criteria:

(a) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change does not involve any physical alteration of

plant configurations, changes to setpoints, or operating parameters.

This proposed amendment is to relocate the T/S requirements for

administrative controls that are related to quality assurance to the

QAPD [Quality Assurance Program Description]. This is in accordance

with the guidance provided in AL 95-06. Also, the relocated

requirements and future changes are controlled by 10 CFR 50.54(a)

which requires prior NRC approval for changes that reduce the

commitments in the program description previously accepted by the

NRC. Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.

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

accident from any previously analyzed.

The proposed change does not involve any physical alteration of

plant configurations, changes to setponts, or operating parameters.

This proposed amendment is to relocate the T/S requirements for

administrative controls that are related to quality assurance to the

QAPD. This is in accordance with the guidance provided in AL 95-06.

Also, the relocated requirements and future changes are controlled

by 10 CFR 50.54(a) which requires prior NRC approval for changes

that reduce the commitments in the program description previously

accepted by the NRC. Therefore, this proposed change does not create

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

previously analyzed.

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

The proposed change does not involve any physical alteration of

plant configurations, changes to setpoints, or operating parameters.

This proposed amendment is to relocate the T/S requirements for

administrative controls that are related to quality assurance to the

QAPD. This is in accordance with the guidance provided in AL 95-06.

Also, the relocated requirements and future changes are controlled

by 10 CFR 50.54(a), which requires prior NRC approval for changes

that reduce the commitments in the program description previously

accepted by the NRC. Therefore, this proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: Mark Reinhart, Acting

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien County, Michigan

Date of amendment request: June 19, 1996 [AEP:NRC:1166AA]

Description of amendment request: The proposed amendment would

modify the technical specifications (T/

[[Page 40023]]

S) to allow continued use of the 2-volt steam generator (SG) tube

plugging criteria for future operating cycles as discussed in NRC

Generic Letter 95-05, ``Voltage-Based Repair Criteria for the Repair of

Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress

Corrosion Cracking.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with the three factor test of 10 CFR 50.92(c),

implementation of the proposed license amendment is analyzed using

the following standards and found not to: 1) involve a significant

increase in the probability or consequences of an accident

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

different kind of accident from any accident previously evaluated;

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

Conformance of the proposed amendment to the standards for a

determination of no significant hazards as defined in 10 CFR 50.92

(three factor test) is shown in the following paragraphs:

1) Operation of Cook Nuclear Plant Unit 1, in accordance with

the proposed license amendment, does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Testing of model boiler specimens for free

span tubing

(no TSP [tube support plate] restraint) at room temperature

conditions show burst pressures in excess of 5000 psi for indications

of outer diameter stress corrosion cracking [ODSCC] with voltage

measurements as high as 19 volts. Burst testing performed on pulled

tubes from Cook Nuclear Plant Unit 1 with up to a 2.02 volt indication

shows measured burst pressure in excess of 10,000 psi at room

temperature. Burst testing performed on pulled tubes from other plants

show burst pressures in excess of 5,300 psi at room temperatures.

Correcting for the effects of temperature on material properties and

minimum strength levels (as the burst testing was done at room

temperature), tube burst resistance significantly exceeds the safety

factor requirements of RG [Regulatory Guide] 1.121 [Bases for Plugging

Degraded PWR Steam Generatory Tubes]. As stated earlier, tube burst

criteria are inherently satisfied during normal operating conditions

due to the proximity of the TSP. Test data indicates that tube burst

cannot occur within the TSP, even for tubes which have 100% throughwall

electric-discharge machined notches 0.75 inch long, provided the TSP is

adjacent to the notched area. Since tube-to-tube support plate

proximity precludes tube burst during normal operating conditions, it

follows that use of the proposed plugging criteria must, therefore,

retain tube integrity characteristics which maintain the RG 1.121

margin of safety of 1.43 times the bounding faulted condition (steam

line break) pressure differential.

During a postulated main SLB [steamline break], the TSP has the

potential to deflect during blowdown, thereby uncovering the

intersection. Based on the existing data base, the RG 1.121

criterion requiring maintenance of a safety factor of 1.43 times the

SLB pressure differential on tube burst is satisfied by 7/8 inch

diameter tubing with bobbin coil indications with signal amplitudes

less than VSL, regardless of the indicated depth measurement. A

2 volt plugging criteria compares favorably with the current

VSL (8.8 volt) structural limit, considering the previously

calculated growth rates for ODSCC within Cook Nuclear Plant Unit 1

SGs. Considering a voltage growth component of 0.8 volts (40%

voltage growth based on 2 volts BOC [beginning of cycle] and a

nondestructive examination uncertainty of 0.40 volts (20% voltage

uncertainty based on 2 volts BOC), when added to the BOC plugging

criteria of 2 volts, results in a bounding EOC [end of cycle]

voltage of approximately 3.2 volts for a cycle operation. A 5.6 volt

safety margin exists (8.8 - 3.2 volt EOC = 5.6 volt margin).

For the voltage/burst correlation, the EOC structural limit is

supported by a voltage of 8.8 volts. Using this VSL of 8.8

volts, a BOC maximum allowable repair limit can be established using

the guidance of RG 1.121. The BOC maximum allowable repair limit

should not permit a significant number of EOC indications to exceed

the VSL and should assure that acceptable tube burst

probabilities are attained. By adding NDE [nondestructive

examination] uncertainty allowances and an allowance for crack

growth to the repair limit, the structural limit can be validated.

The previous plugging criteria submittal established the

conservative NDE uncertainty limit (VNDE) of 20% of the BOC

repair limit. For consistency, a 40% voltage growth allowance

(VGR) to the BOC repair limit is also included. This allowance

is extremely conservative for Cook Nuclear Plant Unit 1. Therefore,

the maximum allowable upper voltage repair limit VURL for BOC,

based on the VSL of 8.8 volts, can be represented by the

expression:

VURL + (VNDE x VURL) + (VGR x VURL) =

8.8 volts, or,

the maximum allowable BOC repair limit can be expressed

as,VURL = 8.8 volt structural limit/1.6 = 5.5 volts.

This structural repair limit supports this application for

plugging criteria implementation to repair bobbin indications

greater than 2 volts based on RPC [rotating pancake coil]

confirmation of the indication. Conservatively, an upper limit of

5.5 volts will be used to repair bobbin coil indications which are

above 2 volts but do not have confirming RPC calls.

Relative to the expected leakage during accident condition

loadings, it has been previously established that a postulated main

SLB outside of containment, but upstream of the main steam isolation

valve, represents the most limiting radiological condition relative

to the plugging criteria. In support of implementation of the

plugging criteria, it will be determined whether the distribution of

crack indications at the TSP intersections at the EOC are projected

to be such that primary-to-secondary leakage would result in site

boundary doses within a small fraction of the 10 CFR 100 guidelines.

A separate calculation has determined this allowable SLB leakage

limit to be 8.4 gpm. Although not required by the Cook Nuclear Plant

design basis, this calculation uses the recommended Iodine-131

transient spiking values consistent with NUREG-0800 [Standard Review

Plan], and the T/S reactor coolant system activity limit of 1 micro

curie per gram dose equivalent Iodine-131. Control room dose

calculations were also performed and found to be less limiting than

the offsite dose leakrate. Therefore, the more conservative offsite

dose leakrate is used. The projected SLB leakage rate calculation

methodology prescribed in GL 95-05 and WCAP 14277 [Steam Line Break

Leak Rate and Tube Burst Probability Analysis Methods for Outside

Diameter Stress Corrosion Cracking at Tube Support Plate

Intersections] will be used to calculate EOC leakage, based on

actual EOC distributions and EOC projected distributions. Due to the

relatively low voltage growth rates at Cook Nuclear Plant Unit 1 and

the relatively small number of indications affected by the plugging

criteria, SLB leakage prediction per GL 95-05 is expected to be

significantly less than the permissible level of 8.4 gpm in the

faulted loop.

The inclusion of all intersections in the leakage model, along

with application of a probability of detection of 0.6, will result

in extremely conservative leakage estimations. Close examination of

the available data shows that indications of less than 2.8 volts

will not be expected to leak during SLB conditions.

The proposed amendment does not result in any increase in the

probability or consequences of an accident previously evaluated

within the cook Nuclear Plant Unit 1 Final Safety Analysis Report

(FSAR).

2) The proposed license amendment does not create the

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

previously evaluated.

Implementation of the proposed SG tube plugging criteria does

not introduce any significant changes to the plant design basis. Use

of the criteria does not provide a mechanism which could result in

an accident outside of the region of the TSP elevations. Neither a

single nor a multiple tube rupture event would, under any plant

conditions, be expected in a SG in which the plugging criteria has

been applied. Specifically, we will continue to implement a maximum

leakage rate limit of 150 gpd (0.1 gpm) per SG to help preclude the

potential for excessive leakage during all plant conditions. The T/S

limits imposed on primary-to-secondary leakage at operating

conditions are a maximum of 0.4 gpm (600 gpd) for all SGs with a

maximum of 150 gpd allowed for any one SG.

The RG 1.121 criteria for establishing operational leakage rate

limits that require

[[Page 40024]]

plant shutdown are based upon leak-before-break (LBB) considerations

to detect a free span crack before potential tube rupture during

faulted plant conditions. The 150 gpd limit should provide for

leakage detection and plant shutdown in the event of the occurrence

of an unexpected single crack resulting in leakage that is

associated with the longest permissible crack length. Regulatory

Guide 1.121 acceptance criteria for establishing operating leakage

limits are based on LBB considerations such that plant shutdown is

initiated if the leakage associated with the longest permissible

crack is exceeded. The longest permissible crack is the length that

provides a factor of safety of 1.43 against bursting at faulted

conditions maximum pressure differential. A voltage amplitude of 8.8

volts for typical ODSCC corresponds to meeting this tube burst

requirement at a lower 95% prediction limit on the burst correlation

coupled with 95/95 lower tolerance limit material properties.

Alternate crack morphologies can correspond to 8.8 volts so that a

unique crack length is not defined by the burst pressure versus

voltage correlation. Consequently, typical burst pressure versus

through-wall crack length correlations were used to define the

``longest permissible crack'' for evaluating operating leakage

limits. Consistent with the cycle 13, 14 and 15 license amendment

requests for plugging criteria, and Section 5 of Enclosure 1 of the

GL, operational leakage limits will remain at 150 gpd per SG. Axial

cracks leaking at this level are expected to provide LBB protection

at both the SLB pressure differential of 2560 psi and, while not

part of any established LBB methodology, LBB protection will also be

provided at a value of 1.43 times the SLB pressure differential.

Thus, the 150 gpd limit provides for plant shutdown prior to

reaching critical crack lengths for SLB conditions. Additionally,

this LBB evaluation assumes that the entire crevice area is

uncovered during blowdown. Partial uncovery will provide benefit to

the burst capacity of the intersection.

3) The proposed license amendment does not involve a significant

reduction in margin of safety.

The use of the voltage-based bobbin probe interim TSP elevation

plugging criteria at Cook Nuclear Plant Unit 1 is demonstrated to

maintain SG tube integrity commensurate with the criteria of RG

1.121. Regulatory Guide 1.121 describes a method acceptable to the

NRC staff for meeting GDC [General Design Criteria] 14, 15, 31, and

32 by reducing the probability or the consequences of SG tube

rupture. This is accomplished by determining the limiting conditions

of degradation of SG tubing, as established by in-service

inspection, for which tubes with unacceptable cracking should be

removed from service. Upon implementation of the criteria, even

under the worst case conditions, the occurrence of ODSCC at the TSP

elevations is not expected to lead to a SG tube rupture event during

normal or faulted plant conditions. It will be confirmed by analysis

and calculation that EOC distribution of crack indications at the

TSP elevations will result in acceptable primary-to-secondary

leakage during all plant conditions and that radiological

consequences are not adversely impacted.

In addressing the combined effects of a LOCA [loss-of-coolant

accident] and SSE [safe-shutdown earthquake] on the SG component (as

required by GDC 2), it has been determined that tube collapse may

occur in the SGs at some plants. The postulated tube collapse

results from a deformation of TSPs as a result of lateral loads at

the wedge supports at the periphery of the plate. The lateral loads

result from the combined effects of the LOCA rarefaction wave and

SSE loadings. The resulting pressure differential on the deformed

tubes may then cause some of the tubes to collapse.

There are two issues associated with a postulated SG tube

collapse. First, the collapse of SG tubing reduces the RCS [reactor

coolant system] 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 peak clad

temperature. Second, there is a potential that partial through-wall

cracks in tubes could progress to through-wall cracks during tube

deformation or collapse.

Consequently, since the LBB methodology is applicable to the

Cook Nuclear Plant Unit 1 reactor coolant loop piping, the

probability of breaks in the primary loop piping is sufficiently low

that they need not be considered in the structural design of the

plant. The limiting LOCA event becomes either the accumulator line

break or the pressurizer surge line break. Loss of coolant accident

loads for the primary pipe breaks were used to bound the Cook

Nuclear Plant Unit 1 smaller breaks. The results of the analysis

using the larger break inputs show that the LOCA loads were found to

be of insufficient magnitude to result in SG tube collapse or

significant deformation.

Addressing RG 1.83 [In-Service Inspection of PWR Steam Generator

Tubes] considerations, implementation of the bobbin coil probe,

voltage-based interim tube plugging criteria of 2 volts is

supplemented by enhanced eddy current inspection guidelines to

provide consistency in voltage normalization, a 100% eddy current

inspection sample size at the TSP elevation per T/S, and MRPC

[motorized RPC] inspection requirements for the larger indications

left in-service to characterize the principal degradation as ODSCC.

As noted previously, implementation of the TSP elevation

plugging criteria will decrease the number of tubes which must be

repaired. The installation of SG tube plugs reduces the RCS flow

margin. Thus, implementation of the plugging criteria will maintain

the margin of flow that would otherwise be reduced in the event of

increased tube plugging.

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

amendment request does not result in a significant reduction in

margin with respect to plant safety as defined in the FSAR or any

Bases of the plant T/Ss.

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: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: Mark Reinhart, Acting

Northern States Power Company, Docket No. 50-282, Prairie Island

Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

Date of amendment request: July 15, 1996

Description of amendment request: The proposed amendment would

allow the use of the moveable incore detector system for measurement of

the core peaking factors with less than 75% and greater than or equal

to 50% of the detector thimbles available. The amendment request is a

one-time only change for Prairie Island, Unit 1, Operating Cycle 18. It

is being submitted to allow for continued operation if the number of

detector thimbles drops below 75%.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes do not involve an increase in the

probability of an accident previously evaluated. The moveable incore

detector system is used only to provide confirmatory information on

the neutron flux distribution and is not required for the daily safe

operation of the core. The system is not a process variable that is

an initial condition in the accident analyses. The only accident

that the moveable incore detector system could be involved in is the

breaching of the detector thimbles which would be enveloped by the

small break loss of coolant accident (LOCA) analysis. As the

proposed changes do not involve any changes to the system's

equipment and no equipment is operated in a new or more harmful

manner, there is no increase in the probability of such an accident.

The proposed amendments would not involve an increase in the

consequences of an accident previously evaluated. The moveable

incore detector system provides a monitoring function that is not

used for accident mitigation (the system is not used in the primary

success path for mitigation of a design basis accident). The ability

of the reactor protection system or engineered

[[Page 40025]]

safety features system instrumentation to mitigate the consequences

of an accident will not be impaired by the proposed changes. The

small break LOCA analysis (and thus its consequences) continues to

bound potential breaching of the system's detector thimbles.

With greater than or equal to 50% and less than 75% of the

detector thimbles available, core peaking factor measurement

uncertainties will be increased, which could impact the core peaking

factors and as a result could affect the consequences of certain

accidents. However, any changes in the core peaking factors

resulting from increased measurement uncertainties will be

compensated for by conservative measurement uncertainty adjustments

in the Technical Specifications to ensure that pertinent core design

parameters are maintained. Sufficient additional penalty is added to

the power distribution measurements such that this change will not

impact the consequences of any accident previously evaluated.

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

proposed changes will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

analyzed.

The proposed amendments would not create the possibility of a

new or different kind of accident previously evaluated as they only

affect the minimum complement of equipment necessary for operability

of the moveable incore detector system. There is no change in plant

configuration, equipment or equipment design. No equipment is

operated in a new manner. Thus the changes will not create any new

or different accident causal mechanisms. The accident analysis in

the Updated Safety Analysis Report remains bounding.

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

proposed changes will not create the possibility of a new or

different kind of accident.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

The proposed changes will not involve a significant reduction in

a margin of safety. The reduction in the minimum complement of

equipment necessary for the operability of the moveable incore

detector system could only impact the monitoring/calibration

functions of the system. Reduction of the number of available

moveable incore detector thimbles to the 50% level does not

significantly degrade the ability of the system to measure core

power distributions. With greater than or equal to 50% and less than

75% of the detector thimbles available, core peaking factor

measurement uncertainties will be increased, but will be compensated

for by conservative measurement uncertainty adjustments in the

Technical Specifications to ensure that pertinent core design

parameters are maintained. Sufficient additional penalty is added to

the power distribution measurements such that this change does not

impact the safety margins which currently exist. Also, the reduction

of available detector thimbles has negligible impact on the quadrant

power tilt and core average axial power shape measurements.

Sufficient detector thimbles will be available to ensure that no

quadrant will be unmonitored.

Based on these factors, the proposed changes in this license

amendment will not result in a significant reduction in the plant's

margin of safety, as the core will continue to be adequately

monitored.

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

Section 50.91, Northern States Power Company has determined that

operation of the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by NRC

regulations in 10 CFR 50, Section 50.92.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: Mark Reinhart, Acting

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 31, 1996

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to add a Limiting Condition

for Operation (LCO) for trisodium phosphate (TSP) and increase the

minimum required amount of TSP contained in the containment sump mesh

baskets.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Trisodium Phosphate Dodecahydrate (TSP) is stored in the

containment sump to raise the pH of the sump and spray water

following a loss of coolant accident (LOCA). As the pH of the water

increases, more radioactive iodine is kept in solution and the

possibility of airborne radioactivity leakage is decreased. An

additional advantage of a higher pH is the beneficial reduction in

chloride stress corrosion cracking (SCC) of austenitic stainless

steel components in the containment following a LOCA.

This chemical is an accident mitigator, not an accident

initiator in that it is not used until after an accident (i.e., a

LOCA) has occurred. At the time it begins to go into solution, the

accident has occurred, containment spray has been activated and

water is collecting in the containment sump. Therefore, increasing

the Technical Specification (TS) minimum amount of TSP verified to

be in containment will not involve a significant increase of the

probability of an accident previously evaluated.

The Updated Safety Analysis Report (USAR), Section 14.15, ``Loss

of Coolant Accident,'' does not take credit for a post-LOCA minimum

containment sump pH adjustment to 7.0 for the iodine removal and

retention calculation until ten hours after initiation of the event.

Increasing the amount of TSP (based on recent re-analysis) in the

containment sump ensures that a pH greater than or equal to 7.0 is

achieved and therefore does not increase the consequences of any

accident previously evaluated.

The proposed change to TS 2.3(4) represents a new Limiting

Condition for Operation (LCO) which is added to establish overall

consistency with the CE STS [Combustion Engineering Standard

Technical Specifications] for TSP requirements. The proposed change

establishes a minimum TSP volume that must be maintained during

operating Modes 1 and 2 to ensure that a pH greater than or equal to

7.0 is achieved within four hours following a LOCA; as well as,

establishing times for accomplishing corrective actions should the

LCO not be met. Therefore, this change does not significantly

increase the probability or consequences of any accident previously

evaluated.

The proposed change to TS 3.6(2)d(i) revises the required

surveillance inventory of the TSP baskets consistent with the

aforementioned calculation to ensure that a pH greater than or equal

to 7.0 is achieved. Therefore, this change does not increase the

consequences of any accident previously evaluated.

The proposed change to TS 3.6(2)d(ii) moves the surveillance

test amounts of chemical and water used from the Specification to

the Basis section. This relocation will not alter the test method or

acceptance criteria.

In the Basis, the amount of TSP used in the test is changed to

reflect the ratio of TSP to water that would be found in the

containment sump following a LOCA. The specified concentration of

boron in the test reflects the highest concentration that could be

found in the containment sump following a LOCA. The test temperature

is changed to 115 - 125 deg.F, which is well below the temperature

expected to be found in the containment sump following a LOCA. The

decanting of the solution does not change the intent of the test

method since the dissolving period will still be conducted without

agitation. Therefore, these changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

[[Page 40026]]

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

or different kind of accident from any accident previously

evaluated.

TSP is currently present in the containment sump. The addition

of TSP ensures that a pH greater than or equal to 7.0 is achieved

following a LOCA. The increase in TSP inventory will be accomplished

via a modification to be installed during the 1996 Refueling Outage.

The proposed change to TS 2.3(4) represents a new LCO which is

added to establish overall consistency with the CE STS for TSP

requirements. The proposed change establishes a minimum TSP volume

that must be maintained during operating Modes 1 and 2 to ensure

that a pH greater than or equal to 7.0 is achieved following a LOCA,

as well as, establishing corrective action term limits should the

LCO not be met. This proposed change does not create a possibility

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

The proposed change to TS 3.6(2)d(ii) moves the surveillance

test amounts of chemical and water used from the Specification to

the Basis section to be consistent with the CE STS. This relocation

will not alter the test method or acceptance criteria. In the Basis

section, the amount of TSP used in the test is changed to reflect

the ratio of TSP to water that would be found in the containment

following a LOCA. The specified concentration of boron in the test

reflects the highest concentration that could be found in the

containment sump following a LOCA. The test temperature is changed

to a range of 115 - 125 deg.F which is well below the temperature

expected to be found in the containment sump following a LOCA. The

decanting of the solution does not change the intent of the test

method since the dissolving period will still be conducted without

agitation. Therefore, these changes will not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

TSP is stored in the containment lower level to raise the pH of

the containment sump and recirculated spray water following a LOCA.

As the pH of the water increases, more radioactive iodine is kept in

solution and the possibility of airborne radioactivity leakage is

decreased. Additionally, a higher pH has the beneficial effect of

reducing the possibility of chloride stress corrosion cracking of

austenitic stainless steel components in the containment.

The proposed change to TS 2.3(4) represents addition of a new

LCO for TSP requirements during power operations and hot standby

consistent with CE STS. This change does not involve a significant

reduction in a margin of safety.

TS 3.6(2)d(i) requires verification that a minimum volume of TSP

is contained in the storage baskets in containment. This change

proposes to increase that volume consistent with the latest ABB/CE

calculation. The increased volume will ensure that the containment

sump, when filled with water from the Reactor Coolant System, Safety

Injection Refueling Water Tank, Safety Injection Tanks and Boric

Acid Storage Tanks, will have a pH greater than or equal to 7.0

within four hours following a LOCA. Therefore, this change does not

involve a reduction in a margin of safety.

The proposed change to TS 3.6(2)d(ii) would move the

surveillance test amounts of chemical and water used from the

Specification to the Basis section. This relocation is consistent

with the CE STS and will not alter the test method or acceptance

criteria. In the Basis, the amount of TSP used in the test is

changed to reflect the ratio of TSP to water that would be found in

the containment following a LOCA. The specified concentration of

boron in the test reflects the highest post-LOCA concentration that

could be found in the containment. The test temperature is changed

to a range of 115 - 125 deg.F which is well below the temperature

expected to be found in the containment sump following a LOCA. The

decanting of the solution does not change the intent of the test

method since the dissolving period will still be conducted without

agitation. Therefore, these 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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: July 15, 1996

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to allow the use of either

zircaloy or ZIRLO cladding and add a reference to Westinghouse Topical

Report, WCAP-12610, June 1990.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed revision to TS 4.3.2 is based on improved STS 4.2

of NUREG-1432. ZIRLO is similar in chemical composition, physical

and mechanical properties to Zircaloy-4, but features improved

corrosion performance and dimensional stability. These

characteristics ensure that fuel rod cladding integrity and fuel

assembly structural integrity are maintained. Fuel assemblies

manufactured with ZIRLO clad fuel rods meet the same design bases

requirements as fuel assemblies manufactured with Zircaloy-4

cladding and the regulatory requirements of 10 CFR 50.46 are

applicable to either material.

No concerns have been identified pertaining to reactor operation

with a core comprised of fuel assemblies manufactured with Zircaloy-

4 clad rods and fuel assemblies manufactured with ZIRLO clad rods.

ZIRLO clad fuel rods do not require a change to the FCS [Fort

Calhoun Station] reload design and safety analysis limits.

Radiological consequences of previously evaluated accidents are not

increased because the safety analysis dose predictions are not

sensitive to the type of cladding material used. The proposed

limited substitution of zirconium alloy or stainless steel filler

rods in accordance with NRC-approved fuel rod configurations will

allow leaking fuel rods (or potential leakers) to be removed.

Therefore, the radiological consequences of accidents previously

evaluated in the FCS Updated Safety Analysis Report (USAR) are not

increased by this change.

The revisions to TS 4.3.2 listed above will not result in a

change to any of the process variables that might initiate an

accident or affect the radiological release for an accident. The

operating limits will not be changed and the analysis methods to

demonstrate operation within the limits will remain in accordance

with NRC-approved methodology. There are no physical changes to the

plant associated with the change to TS 4.3.2 other than the changes

to the fuel assemblies. Therefore, this revision does not involve a

significant increase in the probability or consequences of an

accident previously evaluated because the safety analysis to be

performed for each cycle will continue to demonstrate compliance

with all fuel safety design bases.

The proposed revision of TS 4.3.2 is supported by Westinghouse

Topical Report, WCAP-12610, ``VANTAGE + Fuel Assembly Report,''

dated June 1990 (Westinghouse Proprietary). This topical report

describes the fuel rod design bases, criteria and models, which are

affected by the use of ZIRLO cladding. Consequently, WCAP-12610 is

proposed for addition to the list of analytical methods located in

TS 5.9.5b that are used to determine the core operating limits.

Based on the above discussion, these 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.

Fuel assemblies manufactured with ZIRLO clad fuel rods must meet

original design criteria and thus they will not be an initiator for

any new or different kind of accident. All design and performance

criteria will continue to be met by fuel assemblies manufactured

with ZIRLO clad fuel rods and

[[Page 40027]]

no new single failure mechanisms have been found.

The use of fuel assemblies manufactured with ZIRLO cladding does

not involve any alterations to plant equipment or procedures that

would introduce any new or unique operational modes or accident

precursors. The substitution of zirconium alloy, stainless steel

filler rods, or lead test assemblies for fuel rods will be limited

to NRC-approved fuel rod configurations. Therefore, the possibility

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

evaluated is not created by this change.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The use of fuel assemblies manufactured with ZIRLO clad rods

does not change the proposed FCS reload design and safety analysis

limits. The normal operating conditions allowed for in the Technical

Specifications will be taken into consideration for the use of these

fuel assemblies. For each cycle reload core, the fuel assemblies

will be evaluated using NRC-approved reload design methods to

include consideration of the core physics analysis peaking factors

and core average linear heat rate effects.

NRC-approved methods will also be used to analyze each

configuration of zirconium alloy or stainless steel filler rods in

fuel assemblies to demonstrate continued safe operation within the

limits that assure acceptable plant response to accidents and

transients. Therefore, this change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

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: June 21, 1996

Description of amendment request: The proposed amendment would

change the frequency of instrument channel calibrations in Table 4.1-1,

``Minimum Frequencies for Checks, Calibrations and Test of Instrument

Channels'' to accommodate operation with a 24-month operating cycle.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does operation with the proposed license amendment involve a

significant increase in the probability or consequences of any

accident previously evaluated?

Response:

The proposed changes do not involve a significant increase in

the probability or consequences of any accident previously

evaluated. The proposed changes are being made to extend the

calibration frequency to 24-months for the:

Pressurizer Pressure; Accumulator Level and Pressure; andVolume

Control Tank Level.

These changes are being made, using the guidance of Generic

Letter 91-04, to accommodate a 24-month operating cycle. The

proposed changes in the calibration frequencies do not involve any

plant hardware changes (other than alarm adjustments) or the way the

systems function. The results of the instrumentation drift analysis,

loop accuracy/set point calculations and the evaluation of channel

uncertainties indicate the calibrations can be safely extended to

accommodate the 24-month operating cycle.

The four pressurizer pressure channels are used for high and low

pressure protection (i.e., reactor trip and safety injection) and

for overpower-overtemperature protection. Three of the pressure

channels are also used for pressure control and compensation signals

for rod control. Pressurizer pressure indication is also provided in

the control room for use during normal operation and while using the

EOPs (emergency operating procedure). The loop accuracy/setpoint

calculations confirm that sufficient margin exists between the

pressurizer high and low pressure reactor trip, low pressurizer

pressure SI [safety injection], and overtemperature delta-

temperature analytical limits and the existing field trip settings

based on an extended calibration interval. A small increase in

pressurizer pressure normal indication uncertainty due to increased

sensor drift is within the readability of the indicator and has been

incorporated into the pressurizer pressure initial conditions used

in the evaluation of channel uncertainties (Reference 15) [see

application dated June 21, 1996]. The post-accident indication

uncertainties remain bounded by the existing uncertainties used in

the EOPs. Assurance that the RPS [reactor protection system] and ESF

[engineered safety feature] instrumentation and protection logic

relays will function as required is also provided by on-line

surveillance (channel checks performed each shift and quarterly

channel functional tests) that are designed to detect potential

instrument failures and verify operability of pressurizer pressure

channels.

Water level and pressure in each accumulator is monitored by two

redundant channels designed to provide indication in the control

room. High and low level alarm functions alert the operator to

initiate operations to maintain the accumulator water volume or

pressure within the Technical Specifications limits. The level and

pressure instrumentation do not provide an active protective or

control function and are not required to mitigate an accident

condition. The level (or volume) and pressure limits are important

since they are initial conditions assumed in the safety analysis.

The loop accuracy/setpoint calculations for accumulator level and

pressure were updated to include conservative values for 30-month

calibration uncertainties using Westinghouse sensor drift values and

extrapolated vendor specified uncertainties for rack and indicating

components consistent with industry methods. The increased indicator

uncertainty has been evaluated for both input parameters

(accumulator level and pressure) assumed for the LOCA [loss-of-

coolant accident] and Containment Integrity events (Reference 15)

and a non significant increase in both the peak clad temperature and

containment pressure was identified.

The volume control tank (VCT) level instrumentation is not

required to mitigate the consequences of an accident. The

instrumentation provides control room indication and initiates

automatic actions of the chemical and volume control system (e.g.,

diverts letdown to the holdup tanks on high level, initiates makeup

on low level, changes the charging pump suction on low low level).

The loop accuracy/setpoint calculation for VCT level, updated based

on the increased drift and uncertainty, determined that the existing

setpoints remain valid to ensure the VCT instrumentation can perform

the required design function.

2. Does operation with the proposed license amendment create the

possibility of a new or different kind of accident from any

previously evaluated?

Response:

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

different kind of accident from any previously evaluated. The

proposed changes extend the calibration frequency to 24 months for

the Pressurizer Pressure, Accumulator Pressure and Level, and Volume

Control Tank Level instrumentation to accommodate a 24-month

operating cycle. The proposed changes in calibration frequencies do

not involve any plant hardware changes, nor do they change the way

that the systems function.

The extension of the calibration and surveillance test intervals

were evaluated and the results, documented in Reference 15, indicate

that the calibrations can be safely extended to accommodate the 24-

month operating cycle.

3. Does operation with the proposed license amendment involve a

significant reduction in a margin of safety?

Response:

The proposed changes do not involve a significant reduction in a

margin of safety. The proposed changes extend the calibration

frequency to 24 months for the Pressurizer Pressure, Accumulator

Pressure and Level, and Volume Control Tank Level instrumentation to

accommodate a 24-month operating cycle.

The proposed changes result in an increased instrument channel

uncertainty for the pressurizer pressure. An evaluation (Reference

15) has determined that: all

[[Page 40028]]

current cycle 9 safety analysis limits based on pressurizer pressure

uncertainties remain bounding for extended surveillance intervals

(high and low pressure trips); the safety analysis limits for K1 (a

constant used in the overtemperature [DELTA] T trip setpoint) remain

applicable; and, Engineered Safety Feature Actuation System trip

settings based on pressurizer pressure uncertainty remain bounding

(low pressure safety injection).

The proposed changes result in an increased instrument channel

uncertainty for the accumulator level and pressure. An evaluation

(Reference 15) has determined that increasing the uncertainty

results in non-significant (defined by 10 CFR 50.46(a)(3)(i) as less

than 50 deg.F) increases in the total peak clad temperature (less

than 35 deg.F) for the large break and small break LOCA but the

values remain well within regulatory acceptance criteria. The

evaluation also determined that the peak calculated pressure in

containment following a LOCA would increase due to the lower bound

on pressure and the higher bound on volume in the accumulators. An

assessment of the approximate effect on the peak containment

pressure determined that the Technical Specification integrated leak

rate testing value of 42.42 psig (the licensing basis peak pressure)

remains bounding.

The proposed changes result in an increased instrument channel

uncertainty for the VCT level but there are no changes to any

margins of safety because this instrumentation supports a control

function.

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: Jocelyn A. Mitchell, Acting Director

Southern California Edison Company, et al., Docket No. 50-206, San

Onofre Nuclear Generating Station, Unit No. 1, San Diego County,

California

Date of amendment request: December 22, 1995

Description of amendment request: The proposed change would revise

the San Onofre Unit 1 License Condition to delete a reference to

License Condition 2.C(4) from License Condition 2.D. This change is

being requested to eliminate a reporting requirement for violations of

the physical protection plans that is redundant to reporting

requirements in 10 CFR 73.71 and 10 CFR 73 Appendix G.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility according to this proposed

change involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. The proposed change is considered an administrative change.

It has no impact on the probability or consequences of any of the

accidents previously evaluated. This change revises License

Condition 2.D to remove the burden of duplicate reporting

requirements. This change does not affect the physical protection

program as previously approved by the Nuclear Regulatory Commission

(NRC).

A reporting requirement in License Condition 2.D is being

revised to remove the reference to License Condition 2.C(4) for the

physical protection program. The reporting requirements for the

physical protection program are located in the regulations, 10 CFR

73.71 and 10 CFR 73 Appendix G.

Therefore, the probability and consequences of an accidently

previously evaluated are not affected by these proposed changes.

2. Will operation of the facility according to this proposed

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

from any accident previously evaluated.

No. This proposed change is considered an administrative change.

It has no impact on equipment, systems, or structures such that a

new or different kind of accident is created. This change revises

License Condition 2.D to remove duplicate and unnecessary reporting

requirements for the physical protection program. There is no change

associated with the implementation and maintenance of the physical

protection program as previously approved by the NRC.

Therefore, the possibility of a new or different kind of

accident from an accident previously evaluated is not created.

3. Will operation of the facility according to this proposed

change involve a significant reduction in a margin of safety?

No. This proposed change is considered an administrative change

only. It has no impact on the margin of safety associated with the

physical protection program. This change revises License Condition

2.D to remove duplicative and unnecessary reporting requirements for

the physical protection program. The maintenance and implementation

of the physical protection program is not affected by this change.

Therefore, there will not be a significant reduction in a margin

of safety.

The NRC staff has reviewed the analysis of the licensee 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: Main Library, University of

California, P.O. Box 19557, Irvine, California 92713

Attorney for licensee: James A. Beoletto, Esquire, Southern

California Edison Company, P.O. Box 800, Rosemead, California 91770

NRC Project Director: Seymour H. Weiss

Southern California Edison Company, et al., Docket No. 50-206, San

OnofreNuclear Generating Station, Unit No. 1, San Diego County,

California

Date of amendment request: March 13, 1996

Description of amendment request: The proposed change would revise

San Onofre Unit 1 License Condition 2.D in the Operating (Possession

Only) License to remove a reporting requirement that is redundant to

reporting requirements in 10 CFR 50.72 and 50.73. Additionally, the

proposed change would make administrative and editorial changes in the

Permanently Defueled Technical Specifications, which constitute

Appendix A of the Operating (Possession Only) License.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility according to this proposed

change involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. San Onofre Nuclear Generating Station, Unit 1 (SONGS 1) has

been permanently shut down with its reactor defueled and spent fuel

from the reactor stored in the spent fuel pool. The proposed change

will not modify any of the existing plant configurations, controls,

procedures, or Permanently Defueled Technical Specifications (PDTS)

requirements necessary to assure the integrity and safe operation of

the spent fuel pool.

The requested change to License Condition 2.D will result in not

requiring violations of the PDTS to be reported based on License

Condition 2.D. The basis for this change is that all types of

reportable events applicable to a defueled plant are covered by 10

CFR 50.72 and 50.73, which SONGS 1 is required to implement. Any

other reporting requirements imposed through a license condition are

redundant to reporting requirements contained in 10 CFR 50.72 and

50.73. Therefore, this change is administrative.

The requested changes to the PDTS are also administrative in

nature. They consist of changes to reflect the current nuclear

organization and responsibilities, modify administrative

requirements relating to the Onsite Review Committee, modify a

requirement relating to Final Safety Analysis Report documentation

using NRC guidance, and make editorial corrections and improvements

in the text. Since these changes are administrative, they have no

effect on the accidents previously evaluated.

[[Page 40029]]

Therefore, operation of the facility in accordance with this

proposed change will not involve a significant increase in the

probability or consequences of an accidently previously evaluated.

2. Will operation of the facility according to this proposed

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

from any accident previously evaluated.

No. The proposed changes do not alter the design, configuration,

or method of operation of the plant. The changes to License

Condition 2.D and the PDTS are administrative or editorial.

Therefore, operation of the facility in accordance with this

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Will operation of the facility according to this proposed

change involve a significant reduction in a margin of safety?

No. The proposed changes do not alter the design, configuration,

or method of operation of the plant. Since the proposed changes are

administrative or editorial, the existing plant safety margins are

not reduced.

Therefore, operation of the facility in accordance with this

proposed change will not involve a significant reduction in a margin

of safety.

The NRC staff has reviewed the analysis of the licensee 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: Main Library, University of

California, P.O. Box 19557, Irvine, California 92713

Attorney for licensee: James A. Beoletto, Esquire, Southern

California Edison Company, P.O. Box 800, Rosemead, California 91770

NRC Project Director: Seymour H. Weiss

Southern California Edison Company, et al., Docket Nos. 50-361 and

50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,

San Diego County, California

Date of amendment requests: May 29, 1996

Description of amendment requests: The licensee proposes to revise

improved Technical Specification (TS) 3.5.1, ``Safety Injection Tanks

(SITs),'' to increase the minimum boron concentration in the safety

injection tanks from 1850 parts per million (ppm) to 2200 ppm. This TS

change is being requested to support the planned increase in the

operating cycle length.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Southern California Edison (Edison) is increasing the minimum

boron concentration to maintain the ability of the Safety Injection

Tanks (SITs) to perform their intended safety function consistent

with the increase in fuel enrichment up to 4.8 weight percent (w/o)

Uranium-235 and changing the burnable poison from B4C to Erbia

(Erbium-Oxide Er2O3 and fuel mixture) to increase the

length of the operating cycle. Increasing the minimum boron

concentration in the SITs will maintain the ability of the Emergency

Core Cooling System (ECCS) to control core reactivity during and

following an accident.

No change is being made to the design of the safety injection

system. Consequently, there will be no impact on the probability of

initiating an accident which has been previously evaluated.

Increasing the boron concentration in the SITs will ensure the

ability of this system to mitigate the accidents for which it is

required. No other accident conditions, design conditions, Technical

Specifications, or Technical Specification Bases are affected by

this proposed change in boron concentration.

Therefore, the operation of the facility in accordance with this

proposed change does not involve an 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.

There is no change in plant design or operational methodology

imposed by the increase in SIT boron concentration. This increase in

boron concentration is required because Edison is increasing the

fuel enrichment up to 4.8 w/o Uranium-235 and changing the burnable

poison from B4C to Erbia to achieve a longer cycle length.

Therefore, additional negative reactivity is required at the

beginning of the fuel cycle for these alternate coolant sources.

Edison believes this change in the SIT minimum boron

concentration limit is, in essence, an administrative change. The

SITs are filled from the refueling water storage tank (RWST), which

has a technical specification minimum boron concentration

requirement of 2350 ppm. Edison maintains the RWST boron

concentration higher than the minimum limit. As a result, for the

past several years the SIT boron concentration has been

approximately 2500 ppm, even though the technical specification

lower limit is 1850 ppm. The maximum boron concentration limit is

not being changed. Increasing the SIT minimum boron concentration

limit of the technical specification narrows the existing operating

band, and maintaining the boron concentration between 2200 ppm and

2800 ppm will keep the boron concentration between the current band

of 1850 ppm to 2800 ppm. Therefore, changing the SIT minimum boron

concentration from 1850 ppm to 2200 ppm does not involve a physical

change to the plant.

Therefore, the operation of the facility in accordance with this

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.

With the increase in fuel enrichment up to 4.8 w/o Uranium-235

and changing the burnable poison from B4C to Erbia to increase

the length of the operating cycle, increasing the minimum boron

concentration in the SITs is required to maintain the current

margins of safety.

The calculations were performed to ensure the core remains

subcritical (i.e., conservatively 1% shutdown) with the proposed

boron concentration. In addition to the conservative assumptions

used in the calculation, 50 ppm was added to the results.

Therefore, the operation of the facility in accordance with this

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: Main Library, University of

California, P. O. Box 19557, Irvine, California 92713

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, California 91770

NRC Project Director: William H. Bateman

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: June 12, 1996

Description of amendments request: The proposed amendments would

revise the reactor core safety limits, Overtemperature delta T (OTDT)

and Overpressure delta T (OPDT) reactor trip setpoints and allowable

values, and the power distribution limits associated with

implementation of Relaxed Axial Offset Control (RAOC) and FQ

surveillance. The proposed amendments also include changes to the Bases

associated with these specifications and surveillances.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

[[Page 40030]]

1. The proposed safety limits, reactor trip setpoints, HNF [high

neutron flux] setpoints for MSSVs [main steamline safety valves] out

of service, F[delta]H for LOPAR [low parasitic], and RAOC strategy

changes do not increase the probability or consequences of an

accident previously evaluated in the FSAR [Final Safety Analysis

Report]. The core safety limits and trip setpoints were determined

using the NRC reviewed and approved DNB [departure from nucleate

boiling] methodologies, namely RTDP, and approved DNB correlations.

No new performance requirements are being imposed on any system or

component in order to support the revised core limits. Overall plant

integrity is not reduced. The DNB sensitive transients that are

protected by [OPDT] and [OTDT] were reanalyzed or evaluated. The DNB

design criterion continues to be met. None of these changes directly

initiate an accident; therefore, the probability of an accident has

not increased. No new performance requirements are imposed on any

safety-related equipment. The acceptance criteria for the reanalyses

continue to be met; therefore, the consequences of accidents

previously evaluated in the FSAR are not significantly changed. All

dose consequences have been evaluated for these changes and all

acceptance limits continue to be met. All safety analyses that use

the revised [OTDT] and [OPDT] setpoints continue to meet all

acceptance criteria. [Loss-of-coolant accident] LOCA analyses are

not affected by any of these proposed changes.

2. The proposed Technical Specifications changes do not create

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

accident already evaluated in the FSAR. No new accident scenarios,

failure mechanisms or limiting single failures are introduced as a

result of the proposed changes. The proposed Technical

Specifications changes have no adverse effects on any safety-related

system and do not challenge the performance or integrity of any

safety-related system. The DNB design criterion continues to be met.

The use of the revised core limits, reactor trip setpoints and RAOC

have been shown to allow FNP [Farley Nuclear Plant] to operate in a

safe configuration. Therefore, the possibility of a new or different

kind of accident is not created.

3. The proposed Technical Specifications changes do not involve

a significant reduction in a margin of safety. All accident analysis

acceptance criteria continue to be met. The DNB design criterion

remains unchanged. The DNBR [departure from nucleate boiling ratio]

design limit values have not changed. Therefore, the DNB design

limit values associated with the DNB methodology and correlations,

upon which the Technical Specifications changes are based, do not

result in a significant reduction in the margin of safety because

the DNB design criterion continues to be met. The proposed revisions

to the Technical Specifications result in an operating configuration

consistent with the analytic assumptions (including LOCA analyses)

used to form the bases of the Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: Herbert N. Berkow

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: June 20, 1996

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TS) to incorporate the

requirements of 10 CFR Part 50, Appendix J, Option B. The

Administrative Controls portion would be revised to establish and

reference a ``Containment Leakage Rate Testing Program'' in accordance

with the NRC's Regulatory Guide 1.163 dated September 1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability of consequences of an accident previously evaluated.

The proposed changes provide a mechanism within the TS for

implementing a performance-based leakage rate test program which was

promulgated by the revision to 10 CFR [Part] 50 to incorporate

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

physical or operational changes to structures, systems or

components. The proposed TS Limiting Conditions for Operation (LCO)

are consistent with 10 CFR [Part] 50, Appendix J requirements and

are equivalent to the current LCO requirements. The current safety

analyses and safety design basis for the accident mitigation

functions of the containment, the airlocks, and the containment

isolation valves are maintained. Since the allowable containment

leakage is still maintained within the analyzed limit assumed in the

accident analyses, there is no adverse effect on either onsite or

offsite dose consequences. Furthermore, containment leakage is not

an accident initiator. Therefore, these changes will not increase

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously analyzed.

The proposed changes do not involve any physical or operational

changes to structures, systems or components. No new failure

mechanisms beyond those already considered in the current plant

safety analyses are introduced. Therefore, the proposed changes do

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

from any accident previously analyzed.

3. The proposed changes do not involve a significant reduction

in the margin of safety. Extending Type A, B, and C test intervals

from those currently provided in the TS to those provided for in 10

CFR [Part] 50 Appendix J, Option B slightly increases risk due to an

increased likelihood of containment leakage corresponding to the

increased testing intervals. However, this is somewhat compensated

by the corresponding risk reduction benefits received from the

reduction in component cycling, stress, and wear associated with the

increased intervals. When considering the total integrated risk,

which includes all analyzed accident sequences, the additional risk

associated with increasing test intervals is negligible.

The NRC letter to NEI [Nuclear Energy Institute] dated November

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

FNP [Farley Nuclear Plant] are required to implement Option B of 10

CFR [Part] 50, Appendix J. In NUREG-1493, ``Performance-Based

Containment Leak-Test Program,'' dated September 1995, which forms

the basis for the Appendix J revision, the NRC concludes that

adoption of performance-based test intervals for Appendix J testing

will not significantly reduce the margin of safety. The containment

leak rate data and component performance history at FNP are

consistent with the conclusions reached in NUREG-1493 and NEI 94-01.

Thus, the proposed license amendments do not involve a significant

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

regulatory goal of ensuring an essentially leak-tight containment

boundary.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: Herbert N. Berkow

[[Page 40031]]

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: May 28, 1996

Description of amendment request: The proposed amendment would

increase the test interval for Technical Specification (TS) 3/4.3.1.1,

Reactor Protection System Instrumentation from monthly on a staggered

test basis to semiannually on a staggered test basis for the control

rod drive trip breakers and the reactor trip module logic.

Additionally, the proposed amendment would increase the test interval

from monthly to semiannually for the output logic of the anticipatory

reactor trip system (ARTS) instrumentation as specified in TS 3/

4.3.2.3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

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

presented below: (1)

Operation of the DBNPS in accordance with the proposed license

amendment does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Increasing the surveillance interval will not affect the

probability or consequences of an accident previously evaluated since

performance of the surveillance test only ensures operability of the

particular trip function at the time of the test. The licensee

evaluated the maintenance history and surveillance test results of the

control rod drive trip breakers, reactor trip module logic, and ARTS

output logic to show these components have consistently met their

design and operational requirements over the past 8 years.

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

from any accident previously evaluated.

The proposed changes do not modify or affect system design,

function, operation, or manner of testing.

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

The licensee has performed a reliability evaluation that indicates

insignificant change in reactor trip system unavailability and a

reduction in the potential for spurious trips resulting from testing

which support the conclusion that a significant reduction in a margin

of safety will not occur.

Based on the NRC staff review, it appears that the three standards

of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: June 28, 1996

Description of amendment request: The proposed amendment would

revise the Technical Specifications for shutdown margin to allow

calculational determination of the highest worth control rod. Editorial

changes are also included.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) During refueling, maintenance may be performed on either the

control rods or the control rod drive mechanisms. Controls, such as

refueling interlocks, are provided to assure inadvertent criticality

does not occur during this maintenance. There are no proposed

revisions to these controls except to lower the threshold for

applicability, which constitutes a more restrictive change.

These controls also continue to assure that the new, higher

minimum shutdown margin is maintained to ensure the reactor can be

returned to a subcritical condition should an inadvertent

criticality occur. The proposed alternate calculational method for

highest worth control rod has additional conservatism to account for

any uncertainties in the calculation and provides equivalent margin.

Therefore, this change will not significantly increase the

probability or consequences of any previously analyzed accident.

(2) The proposed change does not necessitate a physical

alteration of the plant in that no new or different type of

equipment will be installed. The proposed change does propose a

higher minimum shutdown margin and a lower threshold of

applicability for CRD [control rod drive] maintenance, both of which

are more restrictive. The proposed change will provide effective

methods to preserve the safety functions associated with the

prevention or automatic mitigation of design basis accidents. Thus,

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

kind of accident from any accident previously evaluated.

(3) The proposed changes to the controls provided to allow

control rod withdrawal for the purposes of maintenance are more

restrictive and thus preserve the safety functions associated with

the prevention or automatic mitigation of design basis accidents.

The addition of a higher minimum shutdown margin requirement and the

proposed calculational alternative for highest worth rod, does not

decrease any of the safety controls or functions to prevent

inadvertent criticalities and provides equivalent or higher margins.

Therefore, this change will not significantly reduce 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Attorney for licensee: R. K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624

NRC Project Director: Jocelyn A. Mitchell, Acting Directorboro, VT

05301

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: July 3, 1996

Description of amendment request: The proposed amendment would

modify Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

Section 4.2.b, ``Steam Generator Tubes,'' to: revise the plugging

criteria for tubes in the tubesheet crevice region; add new inspection

criteria for tubes evaluated using the new plugging criteria; add

definitions of terms used in the new plugging criteria; and add

reporting requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

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

presented below:

1. Operation of the KNPP in accordance with the proposed license

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The revised plugging criteria ensure that tubes in the tubesheet

with indication(s) are sufficiently inspected and evaluated and, if

necessary, rolled to meet the proposed

[[Page 40032]]

acceptance criteria based on the new definitions of acceptable

distance between the indication and the rolled area. With sufficient

distance between the indication(s) and the hard rolled region of the

tube in the tubesheet, tube rupture probability and the consequences

of tube rupture are the same as previously analyzed. Additionally,

the potential for leakage is within previously analyzed limits.

2. The proposed license amendment request does not create the

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

previously evaluated.

Implementation of the proposed tube plugging criteria and

proposed inspection acceptance criteria based on the proposed

definitions does not introduce any significant changes to the plant

design basis. Use of these criteria will not introduce a mechanism

that will result in an accident initiated outside of the tubesheet

crevice region. Any hypothetical accident as a result of tube

indications in the tubesheet crevice region of the tube will be

bounded by the existing tube rupture analysis. Therefore,

application of the revised acceptance criteria for indication(s)

within the tubesheet crevice region will not create the possibility

of a new or different kind of accident.

3. The proposed license amendment does not involve a significant

reduction in the margin of safety.

The use of the proposed inspection criteria and tube plugging

acceptance criteria will maintain the integrity of the tube bundle

commensurate with the requirements of Regulatory Guide 1.121 under

normal and postulated accident conditions. The safety factors used

in verification of the strength of tube(s) evaluated under the new

plugging criteria are consistent with the safety factors in the ASME

Boiler and Pressure Vessel Code used for steam generator design. The

leak testing acceptance criteria are based on the primary-to-

secondary leakage limits in the TSs and the Updated Safety Analysis

Report accident analyses will be maintained. Therefore, the proposed

TS change will not result in a significant reduction in the margin

of safety.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: University of Wisconsin,

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

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

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

NRC Project Director: Gail H. Marcus

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed NoSignificant

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.

Northeast Utilities Service Company, Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London, Connecticut

Date of amendment request: July 3, 1996Brief

Description of amendment request: The proposed amendments would

provide a one-time change to Technical Specification 3.9.1, ``Refueling

Operations, Boron Concentration.'' The proposed change would remove the

requirement that the boron concentration in all filled po

This text is long and has been trimmed here. Open the source document for the complete record.