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

UNITED STATES 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 April 27, 1996, through May 10, 1996. The

last biweekly notice was published on May 8, 1996 (61 FR 20842).

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 June 21, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with

[[Page 25697]]

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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of amendments request: April 5, 1996

Description of amendments request: Pursuant to 10 CFR 50.80 and

50.90, the Baltimore Gas and Electric Company (BGE) hereby requests the

transfer and amendment of Operating License Nos. DPR-53 and DPR-69 for

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2.

The proposed license transfers and amendments are requested as part

of the pending merger between BGE and Potomac Electric Power Company

into Constellation Energy Corporation. The proposed license transfers

would transfer authority to possess and operate Calvert Cliffs from BGE

to Constellation Energy Corporation. The proposed amendments would

change the licenses as well as the related Technical Specifications, to

reflect this transfer by submitting Constellation Energy Corporation in

place of BGE as the licensee for Calvert Cliffs.

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. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed amendment will change the name of the licensee

authorized to possess and operate Calvert Cliffs Nuclear Power Plant

from Baltimore Gas and Electric Company (BGE) to Constellation

Energy Corporation. This amendment request is necessary because of a

proposed merger of BGE and Potomac Electric Power Company into

Constellation Energy Corporation. As a result of the savings

achieved through a reduction in operating costs due to the merger,

Constellation Energy Corporation will have the financial resources

to possess and operate Calvert Cliffs.

In addition, Constellation Energy Corporation personnel will be

technically qualified to operate the plant. Baltimore Gas and

Electric Company nuclear personnel have been named to management

positions in Constellation Energy Corporation, and will remain

responsible for Calvert Cliffs operation and maintenance. The

proposed amendment involves no changes in the training program or

operating organization for Calvert Cliffs.

The proposed amendment does not require any physical change to

the facilities or substantive modifications to the Technical

Specifications or to procedures. The proposed change does not

increase the probability of an accident previously evaluated because

it does not affect any initiators in any previously evaluated

accidents. The proposed change does not increase the consequences of

an accident previously evaluated because it does not affect any of

the items on which the consequences depend.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

of accident from any accident previously evaluated.

The proposed amendment does not modify the plant's configuration

or operations. As a result, no new accident initiators are

introduced. Therefore, the proposed amendment does not create the

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

previously evaluated.

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

safety.

This amendment request is necessary because of a proposed merger

of BGE and Potomac Electric Power Company into Constellation Energy

Corporation. As a result of the savings achieved through a reduction

in operating costs due to the merger, Constellation Energy

Corporation will have the financial resources to possess and operate

Calvert Cliffs. Also, Constellation Energy Corporation personnel

will be technically qualified to operate the plant. Baltimore Gas

and Electric Company nuclear personnel have been named to management

positions in Constellation Energy Corporation, and will remain

responsible for Calvert Cliffs' operation and maintenance. The

proposed amendment involves no changes in the training program or

operating organization for Calvert Cliffs. In addition, the proposed

amendment to substitute Constellation Energy Corporation for BGE

does not result in any changes to the physical design or operation

of the plant. Therefore, the proposed amendment does not involve a

significant reduction in a margin of safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

[[Page 25698]]

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

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

NRC Project Director: Susan Frant Shankman, Acting

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

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

County, North Carolina

Date of amendments request: April 2, 1996

Description of amendments request: The proposed amendments revise

the Brunswick Steam Electric Plant, Units 1 and 2, Technical

Specifications (TS) to allow uprate of the units to 105 percent of

rated thermal power.

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:

. May the proposed activity involve a significant increase in

the probability or consequences of an accident evaluated previously

in the Safety Analysis Report?

The increase in power level, steam flow, feedwater flow and

associated instrument setpoint changes will not significantly

increase the probability or consequences of an accident previously

evaluated.

The probability (frequency of occurrence) of Design Basis

Accidents occurring is not affected by the increase in power level,

as plant equipment will remain in compliance with the applicable

regulatory criteria (ASME Codes, IEEE Standards, NEMA Standards,

Regulatory Guide criteria, etc.). The physical plant changes

necessary to support power uprate include instrument setpoint

changes, indicating meter scale changes for the RWCU [reactor water

cleanup] System flow and Main Steam Flow indicators, Leak Detection,

Process Computer, ERFIS [emergency response facility information

system], and Feedwater System software changes, and SRV [safety/

relief valve] setpoint changes. The setpoints were calculated in

accordance with the CP&L Setpoint Methodology. Utilizing this

methodology ensures scram setpoints (instrument settings that

initiate automatic plant shutdowns) will be established such that

there is no significant increase in scram frequency due to uprate.

No new challenges to safety related equipment will result from power

uprate.

The changes in consequences of hypothetical accidents which

would occur from 102% of the uprated power (2609 MWt), compared to

those previously evaluated from [greater than or equal to] 102% of

the original power (2485 MWt), are not significant, because the

accident evaluations at uprated power will not result in exceeding

the NRC approved acceptance limits. The spectrum of hypothetical

accidents and transients has been investigated, and those accidents/

transients currently evaluated in the UFSAR [Updated Final Safety

Analysis Report] were shown to meet the plant's current regulatory

criteria at uprated conditions (105%). In the area of core design,

for example, the fuel operating limits will still be met at the

uprated power level, and fuel reload analyses show plant transients

will still meet the criteria accepted by the NRC as specified in

NEDO-24011, ``GESTAR II.'' Challenges to fuel or ECCS [emergency

core cooling system] performance have been evaluated and shown to

meet the criteria of 10CFR50 Appendix K. Challenges to the

containment have been evaluated and still meet 10CFR50 Appendix A

Criterion 38, Long Term Cooling, and Criterion 50, Containment.

Bounding events involving radiological releases have been evaluated

and were shown to be well within the criteria of 10CFR100.

2. May the proposed activity create the possibility of a new or

different kind of accident from any accident previously evaluated in

the Safety Analysis Report?

The change in reactor thermal power will not create the

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

previously evaluated.

Equipment that could be affected by power uprate has been

evaluated. No new operating mode, safety related equipment lineup,

accident scenario, or equipment failure mode was identified. The

full spectrum of accident considerations defined in the BNP

[Brunswick Nuclear Plant] UFSAR has been evaluated and no new or

different kind of accident has been identified. Uprate uses

developed technology and applies it within the capabilities of

existing plant equipment in accordance with existing regulatory

criteria including NRC approved codes, standards, and methods.

General Electric has designed BWRs [Boiling Water Reactors] of

higher power levels than the uprated power of any of the currently

uprated BWR/4 fleet and has not identified new power dependent

accidents.

The changes to the Technical Specifications required to

implement power uprate make little change to the plant's

configuration. These changes fall into three major categories. The

first includes those changes resulting from power uprate parameter

changes. These parameter changes, such as the increase in vessel

pressure, temperature and piping system flows are minor in nature.

The evaluations have shown the plant is still within its design

capabilities when operating under these conditions. The changes

required as a result of power uprate will not affect the design

function(s) of currently installed equipment; therefore, there is no

possibility of a new or different kind of failure mode. The second

set of changes is a result of applying setpoint methodology to

calculate TS Allowable Values and Normal Trip Setpoints for

instruments that are directly affected by the parameter changes due

to power uprate. By using CP&L's methodology, the TS values were

calculated to ensure adequate margin exists between the analytical

limit and the TS Allowable Value. The third change include [sic]

setpoints that were reconstituted by the power uprate project.

Again, CP&L methodology was applied and the results show the

setpoints have moved to a more conservative value. This will reduce

the likelihood of spurious scrams and unnecessary challenges to

safety systems while ensuring initiation/actuation equipment

continues to function consistent with existing accident analyses.

3. Does the proposed activity involve a significant reduction in

a margin of safety defined in the basis of any Operating License

Technical Specification?

Power Uprate will not involve a significant reduction in a

margin of safety. The bounding events which had been analyzed in the

UFSAR were reevaluated to demonstrate that power uprate can be

implemented without exceeding any analyzed limit. Because the

applicable safety analysis criteria and limits are satisfied for

power uprate, the margin of safety associated with the safety limits

and other limits identified in the Technical Specifications will be

maintained.

As discussed in Section 5 of GE Nuclear Energy's License Topical

Report NEDO-31984P ``Generic Evaluations of General Electric Boiling

Water Reactor Power Uprate,'' the safety margins prescribed by the

Code of Federal Regulations (CFR) have been maintained by meeting

the appropriate regulatory criteria. Similarly, the margins provided

by the application of the ASME design criteria have been maintained.

The Brunswick unique analysis NEDC-32466P ``Power Uprate Safety

Analysis Report for Brunswick Steam Electric Plant Units 1 and 2''

discusses the effects of power uprate on safety margins for (1) fuel

thermal limits, (2) design basis accidents and the challenges for

fuel, containment and radiological releases, (3) transient analysis,

(4) non-LOCA radiological releases, and (5) environmental

consequences. These evaluations conclude that applicable safety

analysis criteria and limits are satisfied, and thus, the margins of

safety will be maintained.

The changes to the Technical Specification instrumentation will

not involve a reduction in the margin of safety. The calculations

performed for power uprate have established an analytical limit and

calculated the TS Allowable Value and Nominal Trip Setpoint using

formal setpoint methodology. This ensures the instrumentation

functional requirements are met.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

[[Page 25699]]

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: March 29, 1996

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to add an allowance to

complete a TS required surveillance within 24 hours of discovery of a

missed surveillance in accordance with the guidance of Generic Letter

(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical

Specifications (STS) on the Applicability of Limiting Conditions for

Operation and Surveillance Requirements.'' The wording specifying

intervals for testing has been changed to reflect wording consistent

the new STS. Typographical errors in the basis are also being

corrected.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

probability or consequences of an accident previously evaluated.

The proposed changes clarify and incorporates [sic] NRC guidance

for application of extending or moving surveillance intervals by

plus or minus 25%, by elimination of restrictive surveillance

interval descriptions that conflict with NRC guidance, by allowing

for an additional 24 hours to perform missed surveillances, and by

providing a defined finite period for the term ``immediate'' for

Technical Specification (TS) and Inservice Inspection (ISI)

surveillances. The basis for extending or moving surveillances, as

stated in GL 89-14, ``Line-Item Improvements in Technical

Specifications - Removal of the 3.25 Limit on Extending Surveillance

Intervals,'' is to provide plants flexibility for scheduling the

performance of surveillances and to permit consideration of plant

operating conditions that may not be suitable for conducting a

surveillance at the specified time interval. Such operating

conditions include transient plant operation or ongoing surveillance

or maintenance activities. Extending surveillance intervals during

plant operation can result in a benefit to safety when a scheduled

surveillances [sic] is due at a time that is not suitable for

conducting the scheduled surveillance. NUREG-1431, ``Standard

Technical Specifications - Westinghouse Plants,'' states ``the 25%

extension does not significantly degrade the reliability that

results from performing the surveillance at its specified

frequency.'' This is based on the recognition that the most probable

result of any particular surveillance being performed is the

verification of conformance with the surveillance requirements. The

basis for the 24 hour delay period, as stated in the basis for

NUREG-1431, includes consideration of unit conditions, adequate

planning, availability of personnel, the time required to perform

the surveillance, the recognition that the most probable result of

any particular surveillance being performed is the verification of

conformance with the requirements.'' The basis for defining the term

``immediate'' is to provide guidance to plant personnel for

conducting operability testing of the Steam Driven Auxiliary

Feedwater pump after extended shutdown periods in order to minimize

plant risks and not pose an unsafe operational transient during an

unstable plant configuration (i.e., during plant startup). Since

these changes do not affect plant design, operation, or the manner

in which testing is performed, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes clarify and incorporates [sic] NRC guidance

for application of extending or moving surveillance intervals by

plus or minus 25%, by elimination of restrictive surveillance

interval descriptions that conflict with NRC guidance, by allowing

for an additional 24 hours to perform missed surveillances, and by

providing a defined finite period for the term ``immediate'' for TS

and ISI surveillances. Since these changes do not affect plant

design, operation, or the manner in which testing is performed, the

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

kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The changes proposed, with the exception of allowing an

additional 24 hours to complete missed surveillances, are to clarify

existing surveillance intervals and to provide more specific and

detailed criteria without changing current surveillance scheduling

methodologies. The NRC has determined that allowing an additional 24

hours to complete missed surveillance tests minimizes additional

challenges to plant operations such that there is a conservative

balance between the risk associated with performing the surveillance

during stable plant conditions and the risk of imposing a plant

transient due to TS action statements or changing ``modes'' of

operation. These extensions are current industry practices endorsed

by the NRC which provide flexibility for scheduling and performing

surveillances and permit consideration of plant operating conditions

that may not be suitable for conducting a surveillance at either the

specified time interval or inadvertently missing the surveillance

interval. The risk to safety is low in contrast to the alternatives;

therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

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

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

Date of amendment request: April 8, 1996

Description of amendment request: The proposed amendments would

change various sections of the Technical Specifications (TS) to reflect

the transition of fuel supplier from Generic Electric to Siemens Power

Corporation (SPC). The amendments would revise the definitions and

Limiting Conditions for Operation related to Linear Heat Generation

Rate, Critical Power Ratio, Maximum Critical Power Ratio, and Fraction

of Limiting Power Density to incorporate SPC terms and methodology or

to make the TS vendor neutral. Section 6.0 of the TS would be revised

to include SPC references. The proposed amendment also adds a

requirement to adjust the Average Planar Linear Heat Generation Rate

when the reactor is in single loop operation since SPC methodologies

may require this reduction factor for SPC fuel. The SPC methodologies

to be added to the TS have previously been approved by the NRC. The

proposed amendment would also relocate requirements for the traversing

in-core probe system from the TS to the Core Operating Limits Report

and would upgrade the fuel description in Section 5.0 as a line item

from the Improved Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or consequences of

an accident previously evaluated.

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant

[[Page 25700]]

systems designed to mitigate those consequences. Limits will be

established consistent with NRC approved methods to ensure that fuel

performance during normal, transient, and accident conditions is

acceptable. The proposed Technical Specifications amendment reflects

previously approved SPC methodology used to analyze normal

operations, including anticipated operational occurrences (AOOs),

and to determine the potential consequences of accidents.

Licensing Methods and Models

The proposed amendment is to support operation with NRC approved

fuel and licensing methods supplied from Siemens Power Corporation.

In accordance with FSAR Chapter 15, the same accidents and

transients will be analyzed with the new fuel and methods as were

analyzed by GE for GE fuel. The analysis methods and models are NRC

approved (Note the mixed core treatment of critical power ratio is

being addressed under separate correspondence). These approved

methods and models are used to determine the fuel thermal limits.

Traversing In-core Probe (TIP) uncertainty are assumptions in the

approved Siemens core monitoring methodologies. The SPC core

monitoring code enables the site to monitor keff as well as rod

density to perform the reactivity anomaly surveillance. This is

consistent with GE methodology. Therefore, the change in licensing

analysis methods and models does not significantly increase the

probability of an accident or the consequences of an accident

previously identified. The support systems for minimizing the

consequences of transients and accidents are not affected by the

proposed amendment.

New Fuel Design

The use of ATRIUM 9B fuel at LaSalle does not involve a

significant increase in the probability or consequences of any

accident previously evaluated in the FSAR. The ATRIUM-9B fuel is

generically approved for use as a reload BWR fuel type. (See Boiling

Water Reactor Licensing Methodology Summary, Siemens Power

Corporation, EMF-94-217(NP)). Limiting postulated occurrences and

normal operation have been analyzed using NRC-approved methods for

the ATRIUM 9B fuel design to ensure that safety limits are protected

and that acceptable transient and accident performance is

maintained.

The reload fuel has no adverse impact on the performance of in-

core neutron flux instrumentation or control rod drive response. The

ATRIUM-9B fuel design will not adversely affect performance of

neutron instrumentation nor will it adversely affect the movement of

control blades. The exterior dimensions of the ATRIUM-9B fuel

assembly are essentially identical to the GE9B; the ATRIUM-9B fuel

assembly for LaSalle uses a standard fuel channel and normal control

cell positioning (i.e., no offset). Thus, no adverse interactions

with the adjacent control blade and nuclear instrumentation are

anticipated. Additionally, given the above mentioned overall

envelope similarities, no problems are anticipated with other

station equipment such as the fuel storage racks, the new fuel

inspection stand and the spent fuel pool fuel preparation machine.

The ATRIUM 9B design is neutronically compatible with the

existing fuel types and core components in the LaSalle core. SPC

tests have demonstrated that the ATRIUM-9B fuel design is

hydraulically compatible with the GE9 fuel. The bundle pressure drop

characteristics of the ATRIUM 9B bundle are similar to those of the

GE9 fuel design, hence core thermal-hydraulic stability

characteristics are not adversely affected by the ATRIUM 9B design.

An evaluation of the Emergency Procedures is being performed to

ensure that the use of the ATRIUM-9B fuel at LaSalle does not alter

any assumptions previously made in evaluating the radiological

consequences of an accident at LaSalle Station.

Methods approved by the NRC are being used in the evaluation of

fuel performance during normal and abnormal operating conditions.

The ComEd and SPC methods to be used for the cycle specific

transient analyses have been previously NRC approved. The exception

is the mixed core treatment of critical power ratio, which is being

addressed under separate correspondence.

The description of the fuel is expanded to be consistent with

NUREG-1434. The description of the fuel materials, lead test

assembly use, and stating that designs must have been analyzed with

NRC Staff approved codes does not change existing methods; it only

describes them.

Review of the above concludes that the probability of occurrence

and the consequences of an accident previously evaluated in the

safety analysis report have not been significantly increased.

* * * * *

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

from any accident previously evaluated:

Creation of the possibility of a new or different kind of

accident would require the creation of one or more new precursors of

that accident. New accident precursors may be created by

modifications of the plant configuration, including changes in

allowable modes of operation.

Licensing Methods and Models

The proposed Technical Specification amendment reflects

previously approved SPC methodology used to analyze normal

operations, including AOOs, and to determine the potential

consequences of accidents. As stated above, the proposed changes do

not permit modes of reactor operation which differ from those

currently permitted.

New Fuel Design

The basic design concept of a 9x9 fuel pin array with an

internal water box has been used in various lead assembly programs

and in reload quantities in Europe since 1986. WNP-2 has loaded

reload quantities since 1991. Approximately 650 water box assemblies

have been irradiated in the United States through 1995, with a

substantially higher number being irradiated overseas. The NRC has

reviewed and approved the ATRIUM-9B fuel design. (See Boiling Water

Reactor Licensing Methodology Summary, Siemens Power Corporation,

EMF-94-217(NP)). The similarities in fuel design and operation

indicate there would be no expectation of introducing new or

different types of accidents than have been considered for the

existing fuel. Therefore, the use of ATRIUM-9B fuel at LaSalle does

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

from any accident previously evaluated.

* * * * *

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

the following reasons:

The existing margin to safety is provided by the existing

acceptance criteria (e.g., 10CFR50.46 limits). The proposed

Technical Specification amendment reflects previously approved SPC

methodology used to demonstrate that the existing acceptance

criteria are satisfied. The revised methodology has been previously

reviewed and approved by the USNRC for application to reload cores

of GE BWRs. References for the Licensing Topical Reports which

document this methodology, and include the Safety Evaluation Reports

prepared by the USNRC, are added to the Reference section of the

Technical Specifications as part of this amendment.

Licensing Methods and Models

The proposed amendment does not involve changes to the existing

operability criteria. NRC approved methods and established limits

(implemented in the Core Operating Limits Report) ensure acceptable

margin is maintained. The ComEd and SPC reload methodologies for the

ATRIUM-9B reload design are consistent with the Technical

Specification Bases. The Limiting Conditions for Operation are taken

into consideration while performing the cycle specific and generic

reload safety analyses. NRC approved methods are listed in

Specification 6.0 of the Technical Specifications.

Analyses performed with NRC-approved methodology have

demonstrated that fuel design and licensing criteria will be met

during normal and abnormal operating conditions. Therefore, there is

not a significant reduction in the margin of safety.

New Fuel Design

The exterior dimensions of the ATRIUM-9B fuel assembly are

essentially identical to the GE9B; the ATRIUM-9B fuel assembly for

LaSalle uses a standard fuel channel and normal control cell

positioning; i.e., no offset. Thus, no adverse interactions with the

adjacent control blade and nuclear instrumentation are anticipated.

The change does not adversely impact equipment important to safety

and, therefore does not reduce the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One

[[Page 25701]]

First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: April 9, 1996

Description of amendment request: The proposed amendments would

eliminate the automatic reactor scram function and the group 1 and 3

isolation valve closure functions associated with the Main Steam Line

Radiation Monitoring (MSLRM) system high radiation setpoint.

Elimination of these functions will eliminate potential spurious scrams

and isolations caused by increased main steam line radiation levels

during hydrogen injection. The licensee also proposes to raise the

MSLRM system alarm setpoints which are not part of the Technical

Specifications to include increased background radiation during

hydrogen injection. The proposed amendment would also delete the

surveillance requirements for the associated instruments.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

Redefining the full power radiation background, thus changing

the MSLRM alarm setpoint, does not change the probability of

occurrence of any accident which has been postulated and analyzed in

the UFSAR, but will reduce the probability of the inadvertent MSIV

closure transient which is an analyzed transient in the UFSAR. It

does not change the probability of malfunction of any equipment

important to safety associated with [loss of coolant accident] LOCA,

fuel handling accident or [control rod drop accident] CRDA. It also

does not change the resultant offsite radiological dose from the

bounding design basis CRDA. This is based upon all radioactivity,

resulting from the design basis CRDA, going to the condenser

instantaneously (or independent of the actual MSLRM setpoint) in the

offsite dose calculation.

The elimination of reactor scram and isolation of MSIVs,

isolation of main steam line drain valves and reactor water sample

line valves, associated with the MSLRM system actuation do not

introduce, mitigate, or reduce the probability of any design basis

accident, or any accident, evaluated in the UFSAR. The topical

report NEDO-31400A has shown that there is essentially no reasonable

radiological consequence benefit in a design basis CRDA of retaining

the MSLRM associated reactor scram and MSIV isolation function. In

addition, the probability of inadvertent scram and isolation is

reduced. The proposed change will not adversely impact the operation

of the [reactor protection system] RPS or [primary containment

isolation system] PCIS with respect to performing its other intended

safety functions. The proposed change will not affect the operation

of other plant systems or equipment important to safety. The

consequences of eliminating the automatic closure of the main steam

line drain isolation valves and reactor recirculation water sample

line isolation valves along with the MSIVs has been evaluated to be

negligible additions to the CRDA doses. A [LaSalle County Station]

LSCS unique analysis has demonstrated that the radiological doses as

a result of design basis CRDA are acceptable.

The MSLRM system high radiation trip was intended to function in

response to a CRDA which has been previously evaluated. No credit

for MSIV closure was taken in the CRDA analysis since it postulates

that all the radioactive material assumed to be released from the

fuel is transported to the main condenser prior to MSIV closure.

Furthermore, the probability of a fuel failure is independent of the

operation of the MSLRM system.

By eliminating the MSLRM induced MSIV closure, the Offgas system

can be utilized to reduce potential offsite doses after a CRDA. The

[mechanical vacuum pump] MVP is tripped no later than 15 minutes of

a Hi-Hi radiation alarm but analytically results in acceptable

offsite doses.

Thus the proposed amendment will not increase the probability of

any accident previously evaluated, and the elimination of the MSLRM

isolation signal for MSIVs and other small containment valves will

not significantly increase the consequences of a CRDA as previously

evaluated.

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

from any accident previously evaluated because:

Redefining the full power radiation background, thus changing

the actual MSLRM alarm setpoint, does not alter the configuration of

the plant. It does not revise any logic or function of the MSLRM

trip channels or add, replace, or delete any equipment important to

safety. Therefore it does not introduce any new failure modes or

create any possibility of a new accident which may challenge safety

to the public and has not been previously analyzed. It also does not

involve any equipment which either has not been evaluated

previously, or may have any safety consequences to the public.

The proposed Technical Specification changes involve eliminating

the MSLRM system high radiation trip function for initiating an

automatic reactor scram, and automatic isolations. The proposed

changes will not affect the operation of other plant systems or

equipment important to safety. The MSLRM system will continue to

initiate alarms as before. Plant procedures will be in place to take

appropriate mitigative measures in response to a high alarm.

The isolation and reactor scram functions associated with the

MSLRM system actuation were originally intended to mitigate, not

prevent, a potential accident scenario such as a CRDA or gross fuel

failure event. Adding or removing an electronic signal, such as the

one from the MSLRM system, does not change system or hardware design

within the reactor vessel pressure boundary, and therefore will not

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

those evaluated in the UFSAR like a LOCA or CRDA during power

operation. It also does not create the possibility of a new or

different kind of accident outside the reactor vessel pressure

boundary from those evaluated in the UFSAR, such as a LOCA or Fuel

Handling Accident. Removing the isolation signal also reduces the

probability of inadvertent scram and isolation.

Therefore the proposed amendment will not create the possibility

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

analyzed.

3) Involve a significant reduction in the margin of safety

because:

The current MSLRM trip Hi-Hi alarm setpoint (about 4 R/hour with

full power background at 1.3 R/hour) is at 3 times the full power

radiation background. As indicated in the plant unique analytical

result for LSCS, the radiological reading at the MSLRMs for design

basis CRDA is equivalent to over 1200 times the normal full power

radiation background (1600 R/hour divided by 1.3 R/hour), or 150

times the full power radiation background during peak HWC

environment (since the radiation background is 8 times the normal

background). Thus the safety margin was very large, and would still

be quite large with the HWC background factored into the MSLRM

actuation setpoint (3 x 8 x 1.3 = about 50). The Hi alarm setpoint

of 1.5 times full power background likewise will have a higher

safety margin. Thus there is basically no adverse consequence to the

margin of safety in the basis for the LaSalle technical

specifications.

The proposed Technical Specification changes to eliminate the

MSLRM system high radiation trip function for initiating an

automatic reactor scram, and automatic closure of the MSIVs, main

steam line drain isolation valves, and reactor recirculation water

sample line isolation valves do not cause radiological dose

consequences to exceed the limit established by SRP 15.4.9.

Per NEDO-31400A, the elimination of MSLRM trip/scram signal will

result in the reduction of potential inadvertent scrams, unnecessary

safety-related actuations, undue vessel isolation, and duty

challenges during normal plant operation. These can be interpreted

to be a potential reduction in core damage frequency, which

translates to an improvement in the margin of safety.

Thus the margin of safety as defined in the basis of the

technical specifications is essentially unaffected, and is therefore

acceptable.

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

[[Page 25702]]

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: April 16, 1996

Description of amendment request: The proposed amendments would

eliminate the Technical Specification requirement to perform response

time testing for selected instruments. The instruments affected are the

sensors for selected reactor protection system instrumentation, main

steam isolation actuation instrumentation, and all sensors for

emergency core cooling system (ECCS) actuation instrumentation. The

proposed changes are supported by analyses performed by the Boiling

Water Reactor Owners' Group as documented in NEDO-32291-A which was

approved by the NRC for use in license amendment applications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Involve a significant increase in the probability or consequences of

an accident previously evaluated because:

The purpose of the proposed Technical Specification (TS) change

is to eliminate response time testing requirements for selected

components in the Reactor Protection System (RPS), Isolation

Actuation instrumentation and Emergency Core Cooling System (ECCS)

actuation instrumentation. The Boiling Water Reactor Owners' Group

(BWROG) has completed an evaluation which demonstrates that response

time testing is redundant to the other TS-required testing. These

other tests, in conjunction with actions taken in response to NRC

Bulletin 90-01, ``Loss of Fill-Oil in Transmitters Manufactured by

Rosemount,'' and Supplement 1, are sufficient to identify failure

modes or degradations in instrument response time and ensure

operation of the associated systems within acceptable limits. There

are no known failure modes that can be detected by response time

testing that cannot also be detected by the other TS-required

testing. This evaluation was documented in NEDO-32291-A, ``System

Analyses for the Elimination of Selected Response Time Testing

Requirements,'' dated October 1995. LaSalle County Station, LaSalle,

has confirmed the applicability of this evaluation to LaSalle. In

addition, LaSalle will complete the actions identified in the NRC

staffs safety evaluation of NEDO-32291-A.

Because of the continued application of other existing TS-

required tests such as channel calibrations, channel checks, channel

functional tests, and logic system functional tests, the response

time of these systems will be maintained within the acceptance

limits assumed in plant safety analyses and required for successful

mitigation of an initiating event. The proposed changes do not

affect the capability of the associated systems to perform their

intended function within their required response time, nor do the

proposed changes themselves affect the operation of any equipment.

As a result, LaSalle has concluded that the proposed changes do not

involve a significant increase in the probability or the

consequences of an accident previously evaluated.

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

from any accident previously evaluated because:

The proposed changes only apply to the testing requirements for

the components identified above and do not result in any physical

change to these or other components or their operation. As a result

no new failure modes are introduced. Therefore, the proposed changes

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

from any accident previously evaluated.

3) Involve a significant reduction in the margin of safety

because:

The current TS-required response times are based on the maximum

allowable values assumed in the plant safety analyses. These

analyses conservatively establish the margin of safety. As described

above, the proposed changes do not affect the capability of the

associated systems to perform their intended function within the

allowed response time used as the basis for the plant safety

analyses. The potential failure modes for the components within the

scope of this request were evaluated for impact on instrument

response time. This evaluation confirmed that, with the exception of

loss of fill-oil of Rosemount transmitters, the remaining TS-

required testing is sufficient to identify failure modes or

degradations in instrument response times and ensure that operation

of the applicable instrumentation is within acceptable limits. The

actions taken in response to NRC Bulletin 90-01 and Supplement 1 are

adequate to identify loss of fill-oil failures of Rosemount

transmitters. As a result, it has been concluded that plant and

system response to an initiating event will remain in compliance

with the assumptions of the safety analysis.

Further, although not explicitly evaluated, the proposed changes

will provide an improvement to plant safety and operation by the

following:

a. Reducing the time safety systems are unavailable,

b. Reducing the potential for safety system actuations,

c. Reducing plant shutdown risk,

d. Limiting radiation exposure to plant personnel, and

e. Eliminating the diversion of key personnel resources to

conduct unnecessary testing.

Therefore, LaSalle has concluded that this request will not

significantly reduce the margin of safety, and may actually cause an

increase in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: November 2, 1994

Description of amendment request: The proposed amendments would

delete the content of Appendix B, ``Environmental Protection Plan''

(nonradiological), and modify License Condition 2.C.(2) to delete that

portion which refers to the Environmental Protection Plan.

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 amendments would not involve a significant

increase in the probability or consequences of an accident

previously evaluated]:

Deletion of the Environmental Protection Plan and modifying

License Condition 2.C.(2) will have no impact on the probability or

consequences of an accident previously evaluated because the changes

will not have any impact upon the design or operation of any plant

systems or components.

2. [The proposed amendments would not create the possibility of

a new or different kind of accident from any accident previously

evaluated]:

The proposed revision will not create the possibility of a new

or different kind of accident from any previously evaluated because

the revision is administrative in nature and will not change the

types and amounts of effluent that will be released.

3. [The proposed amendments would not involve a significant

reduction in a margin of safety]:

[[Page 25703]]

The proposed revision will not reduce a margin of safety because

it is administrative in nature and will not [a]ffect the margin of

safety as defined in the basis for any 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley

Power Station, Unit 2, Shippingport, Pennsylvania

Date of amendment request: April 29, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 5.3.1 to allow the use of ZIRCO as

an alternate zirconium-based fuel rod material and remove the word clad

since it has been eliminated from the text of the NRC's improved

Standard Technical Specifications (NUREG-1431). Limited substitution of

fuel rods by ZIRCO filler rods would also be permitted. The proposed

amendment would revise Note 2 on TS Table 3.9-1 to specify that the

maximum burnup in the peak fuel rod in a fuel assembly stored in Region

2 spent fuel racks should not exceed the NRC-approved limit for WCAP-

12610 rather than the current maximum burnup limit of 60 GWD/MTU.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The methodologies used in the accident analyses remain

unchanged. The proposed changes do not change or alter the design

assumptions for the systems or components used to mitigate the

consequences of an accident. Use of ZIRLO fuel rod material does not

adversely affect fuel performance or impact nuclear design

methodology. Therefore, accident analysis results are not impacted.

The operating limits will not be changed and the analysis

methods to demonstrate operation within the limits will remain in

accordance with NRC approved methodologies. Other than the changes

to the fuel assemblies, there are no physical changes to the plant

associated with this technical specification change. A safety

analysis will continue to be performed for each cycle to demonstrate

compliance with all fuel safety design bases.

VANTAGE 5 fuel assemblies with ZIRLO fuel rods meet the same

fuel assembly and fuel rod design bases as other VANTAGE 5 fuel

assemblies. In addition, the 10 CFR 50.46 criteria are applied to

the ZIRLO fuel rods. The use of these fuel assemblies will not

result in a change to the reload design and safety analysis limits.

Since the original design criteria are met, the ZIRLO fuel rods will

not be an initiator for any new accident. The fuel rod material is

similar in chemical composition and has similar physical and

mechanical properties as Zircaloy-4. Thus, the fuel rod integrity is

maintained and the structural integrity of the fuel assembly is not

affected. ZIRLO improves corrosion performance and dimensional

stability. No concerns have been identified with respect to the use

of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel

rods.

The dose predictions in the safety analyses are not sensitive to

the fuel rod material used; therefore, the radiological consequences

of accidents previously evaluated in the safety analysis remain

valid. A reload analysis is completed for each cycle, in accordance

with NRC approved methodologies. Therefore, the proposed change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated.

VANTAGE 5 fuel assemblies with ZIRLO fuel rods satisfy the same

design bases as those used for other VANTAGE 5 fuel assemblies. All

design and performance criteria continue to be met and no new

failure mechanisms have been identified. The ZIRLO fuel rod material

offers improved corrosion resistance and structural integrity.

The proposed changes do not affect the design or operation of

any system or component in the plant. The safety functions of the

related structures, systems, or components are not changed in any

manner, nor is the reliability of any structure, system, or

component reduced. The changes do not affect the manner by which the

facility is operated and do not change any facility design feature,

structure, or system. No new or different type of equipment will be

installed. Since there is no change to the facility or operating

procedures, and the safety functions and reliability of structures,

systems, or components are not affected, the proposed changes do 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 Zircaloy-4, ZIRLO, or stainless steal filler rods in

fuel assemblies will not involve a significant reduction in the

margin of safety because analyses using NRC approved methodology

will be performed for each configuration to demonstrate continued

operation within the limits that assure acceptable plant response to

accidents and transients. These analyses will be performed using NRC

approved methods that have been approved for application to the fuel

configuration.

Use of ZIRLO as fuel rod material does not change the VANTAGE 5

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

assemblies will take into consideration the normal core operating

conditions allowed in the technical specifications. For each reload

core, the fuel assemblies will be evaluated using NRC approved

reload design methods, including consideration of the core physics

analysis peaking factors and core average linear heat rate effects.

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 [Updated

Final Safety Analysis Report] or any plant technical specification

BASES.

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, Pennsylvania 1500l.

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 Nos. 50-313 and 50-368, Arkansas

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

Date of amendment request: May 2, 1996

Description of amendment request: The proposed technical

specification amendments would extend the allowed outage times for

emergency diesel generators at Arkansas Nuclear One, Units 1 and 2 to 7

days with an additional, once per refueling cycle extension of 7 more

days for each machine.

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 emergency diesel generators (EDGs) are backup alternating

current power sources

[[Page 25704]]

designed to power essential safety systems in the event of a loss of

offsite power. The EDGs are not accident initiators in any accident

previously evaluated. Probabilistic Safety Analysis (PSA) methods

were utilized in order to fully evaluate the EDG allowed outage time

(AOT) extension proposed in this submittal. The results of these

analyses indicate there is not a significant increase in the

probability of an accident previously evaluated. Therefore, this

change does not involve an increase in the probability of an

accident previously evaluated.

The EDGs provide backup power to components that mitigate the

consequences of accidents. The current TSs allow for an EDG to be

removed from service for an AOT. The proposed amendment extends the

current AOT for an EDG. The proposed change does not allow any more

equipment to be removed from service at one time. The proposed

changes to the AOTs do not affect any of the assumptions used in

deterministic safety analysis. By extending the EDG AOT, the

consequences of an accident previously evaluated will remain

unchanged.

The proposed change removes redundant requirements associated

with an inoperable emergency power supply from the TS for the

pressurizer proportional heaters. The operability requirements for

emergency power supplies and actions to be taken if an EDG is

inoperable are already addressed in the ANO-2 TS 3.8.1.1.

The associated changes that remove the requirements to test the

EDGs if one or both offsite power supplies are inoperable, for an

inoperable station battery, for an inoperable component in the two

ESF electrical distribution systems, the accelerated testing

requirements of the EDGs, and the daily testing requirements for the

operable EDGs improve the reliability for the operable EDGs by

reducing the number of unnecessary starts and stops. By improving

the EDG reliability, this change will not increase the consequences

of the accidents previously evaluated.

The other changes in this submittal associated with the bases

are considered administrative in nature and have no effect on the

consequences of an accident previously evaluated.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

This proposed change does not alter 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.

The proposed changes do not affect the Technical Specification

limiting conditions for operation or their bases which support the

deterministic analyses used to establish the margin of safety.

Calculations performed to analyze the change in risk based on

these changes produced acceptable values which are included in the

tables located in the description of changes section. These

calculated changes in risk fall well within that which is normally

considered acceptable. When the additional benefit of maintaining

the Emergency Diesel Generators available during shutdown cooling

operations associated with refueling outages in considered, the

overall change in risk is further reduced.

The remaining proposed changes are either associated with

increasing EDG reliability or considered administrative in nature.

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., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: November 20, 1995, as supplemented by

the letter dated December 15, 1995.

Description of amendment request: The licensee has proposed to

revise the Grand Gulf Nuclear Station (GGNS), Unit 1, Technical

Specifications (TSs) as follows for the drywell, the drywell airlock,

and the drywell isolation valves:

1. For the drywell in Limiting Condition of Operation (LCO)

3.6.5.1, the surveillance frequency interval for the drywell bypass

test in Surveillance Requirement (SR) 3.6.5.1.1 would be increased from

18 months to 10 years. For this interval change, an increased testing

frequency would be required if bypass performance degrades (i.e., the

leakage is greater than the limit for two consecutive tests) and the

application of SR 3.0.2, the allowance to extend the surveillance

interval by 25 percent, would be restricted to 12 months on the 10-year

interval. This includes deleting the Note in SR 3.6.5.1.1.

2. For the drywell airlock in LCO 3.6.5.2, the following changes

are requested: (a) the leak rate SR 3.6.5.2.2 would be transferred from

the airlock LCO (3.6.5.2) to SR 3.6.5.1.3 in the drywell LCO (3.6.5.1),

(b) the requirement in SR 3.6.5.2.2 for the air lock to meet a specific

overall leakage limit would be deleted, (c) the Note in SR 3.6.5.2.2

that stated that an inoperable air lock door does not invalidate the

previous air lock leakage test would be deleted, (d) the test pressure

for the air lock leakage test in SR 3.6.5.2.2 would be reduced from

11.5 psig to 3 psid, and (e) the surveillance frequency interval for

the air lock leakage and interlock testing, required in SRs 3.6.5.2.1

and 3.6.5.2.2, would be increased from 18 months to 24 months.

3. For the drywell airlock in LCO 3.6.5.2 and the drywell isolation

valves in LCO 3.6.5.3, the Action Notes, which identify that the

actions required by drywell LCO 3.6.5.1 must be taken when the drywell

bypass leakage limit is not met, would be deleted. Action C.1 of LCO

3.6.5.2 and its associated completion time would also be deleted.There

would also be changes to the Bases of the TSs for the above LCOs and

SRs, based on the proposed changes.

Basis for proposed no significant hazards consideration

determination: The amendment request dated November 20, 1995, applied

to both the Grand Gulf Nuclear Station (GGNS) and the River Bend

Station (RSB); however, not all of the proposed amendments apply to

GGNS. This Notice only discusses the amendment request for GGNS. The

reference below to proposed amendments which do not apply to GGNS are

marked by ``[....]''.

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

analysis of the issue of no significant hazards consideration in its

application dated November 20, 1995, which is presented below:

Entergy Operations, Inc. proposes to change the current Grand

Gulf Nuclear Station (GGNS) [....] Technical Specifications. The

specific proposed changes are:

1. The Surveillance Frequency [interval] for the drywell bypass

test is changed [increased] from 18 months to 10 years with an

increased testing frequency required if performance degrades.

2. The following changes are requested for the drywell air lock

testing: (a) the leakage rate surveillance is moved from the air

lock Limiting Condition for Operation (LCO) to the drywell LCO, (b)

the requirement for the air lock to meet a specific overall leakage

limit is deleted, (c) the Note that an inoperable air lock door does

not invalidate the previous air lock leakage test is deleted, (d)

the GGNS test pressure for the air lock leakage test is changed

[reduced] from 11.5 psig to 3 psid, [...,] and ([e]) the

Surveillance Frequency [interval] for the air lock leakage test and

interlock test is changed [increased] from 18 months to 24 months.

3. The Actions Notes in the drywell air lock LCO and the drywell

isolation valve LCO that identifies that the Actions required

[[Page 25705]]

by the drywell LCO must be taken when the drywell bypass leakage

limit is not met is deleted. [Action C.1 of LCO 3.6.5.2 and its

associated completion time would also be deleted.]

[4. ...]

The Commission has provided standards for determining whether a

no significant hazards consideration exists as stated in 10 CFR

50.92(c). The proposed changes involve the withdrawal of operating

restrictions previously imposed because acceptable operation of the

Mark III primary containment design had not been demonstrated at the

time of licensing. As published in the Federal Register regarding no

significant hazards consideration criteria, granting of a relief,

based upon demonstration of acceptable operation from an operating

restriction that was imposed because acceptable operation had not

yet been demonstrated does not involve a significant hazards

consideration (Ref. 48 FR 14870). Furthermore, a proposed amendment

to an operating license involves no significant hazards

consideration if operation of the facility in accordance with the

proposed amendment would not: (1) involve a significant increase in

the probability or consequences of an accident previously evaluated;

or (2) create the possibility of a new or different kind of accident

from any accident previously evaluated; or (3) involve a significant

reduction in a margin of safety.

Entergy Operations, Inc. has evaluated the no significant

hazards consideration in its request for this license amendment,

even though the above-mentioned criterion is satisfied by this

proposal. In accordance with 10 CFR 50.91(a), Entergy Operations,

Inc. is providing the analysis of the proposed amendment against the

three standards in 10 CFR 50.92(c). A description of the no

significant hazards consideration determination follows:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The requested changes are either administrative changes which

clarify the format of the requirement or change the requirement to

match the design bases of the plant, a change which relocates the

requirement to the Technical Specification Bases, or a change in

[the] surveillance interval. Each of these types of change are

discussed below:

1. The administrative changes clarify the format of the

requirement or change the requirement to match the design bases of

the plant. Clarifying [the] administrative format of the Technical

Specifications does not result in any changes to the Technical

Specification requirements and, as a result, does not involve a

significant increase in the probability or consequences of an

accident previously evaluated. Also, changing the requirements of

the Technical Specifications to more closely match the design bases

of the plant will continue to assure that the plant will respond as

assumed in the accident analyses and, as a result, does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

2. The proposed changes relocate information to the Technical

Specification Bases. In the Technical Specifications Bases the

relocated information will be maintained in accordance with 10 CFR

50.59 and subject to the change control provisions in Chapter 5 of

Technical Specifications. Since any changes to the Technical

Specifications Bases will be evaluated per the requirements of 10

CFR 50.59, no increase (significant or insignificant) in the

probability or consequences of an accident previously evaluated will

be allowed. Therefore, this change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

3. The proposed changes in frequency for the drywell bypass

leakage and drywell air lock surveillances will continue to ensure

that no paths exist through passive drywell boundary components that

would permit gross leakage from the drywell to the primary

containment air space and result in bypassing the primary

containment pressure-suppression feature beyond the design basis

limit. The Mark III primary containment system satisfies General

Design Criterion 16 of Appendix A to 10 CFR Part 50. Maximum drywell

bypass leakage was determined previously by reviewing the full range

of postulated primary system break sizes. The limiting case was a

primary system small break loss of coolant accident (LOCA) and

yielded a design allowable drywell bypass leakage rate limit of

approximately 35,000 scfm for GGNS [....]. The Technical

Specifications acceptable limit for the bypass leakage following a

surveillance is less than 10% of this design basis value. The most

recent bypass leakage value was approximately 2.5% for GGNS [....]

of the design allowable leakage rate limit for the limiting event.

EOI [Entergy Operations, Inc.] is committed to maintaining

programmatic and oversight controls that ensure that drywell bypass

leakage remains a small fraction of the design allowable leakage

limit.

The drywell is typically exposed to essentially 0 psig during

normal plant operation and 3 psig during drywell bypass leak rate

testing. These pressures are considerably lower than the structural

integrity test pressure and are less likely to initiate a crack or

cause an existing crack to grow. Visual inspections of the

accessible drywell surfaces that have been performed since the

structural integrity tests have not revealed the presence of

additional cracking or other abnormalities. Therefore, additional

cracking of the drywell structure is not expected due to testing or

operation and, similar to the justification for the ten year 10 CFR

50 Appendix J Type A test interval, it is not considered credible

for the passive drywell structure to begin to leak sufficiently to

impact the design drywell bypass leakage limit.

The primary containment's ability to perform its safety function

is fairly insensitive to the amount of drywell leakage, thereby

providing a margin to loss of the drywell safety function that is

not normally available for systems. This insensitivity is

demonstrated by the extremely high limiting event design basis

allowable leakage for the drywell (e.g., 35,000 scfm for GGNS

[....]). The limiting leakage is almost an order of magnitude higher

for other events. Additionally, an even higher allowable leakage can

be realistically accommodated by the primary containment due to the

margins in the containment design. Because of the margins available,

it will take valves in multiple penetration flow paths leaking

excessively to cause the primary containment to fail as a result of

overpressurization, the probability that drywell isolation valve

leakage will result in primary containment failure due to excessive

drywell leakage is not considered significant and this drywell/

primary containment failure mode is not considered credible.

The proposed Technical Specification changes have no significant

impact on the GGNS Individual Plant Examination (IPE) [....]

conducted per NRC Generic Letter 88-20. The IPEs considered

overpressurization failure of primary containment as part of the

primary containment performance assessment. Due to the magnitude of

acceptable drywell leakage and the extremely low probabilities of

achieving such leakage, primary containment failure due to

preexisting excessive drywell leakage was considered a non

significant contributor to primary containment failure. Primary

containment overpressurization failure can occur with or without

preexisting excessive drywell leakage in a severe accident. This is

due to physical phenomena associated with potentially extreme

environmental conditions inside primary containment following a

severe accident. However, the calculated frequency of such extreme

conditions is very small. The proposed changes do not impact the IPE

evaluated phenomena causing primary containment overpressurization

failure nor significantly increase the probability that the drywell

has preexisting excessive leakage and therefore would not contribute

to these accident scenarios.

For the reasons discussed above, the proposed changes do not

have any significant risk impact to accidents previously evaluated

and do not significantly increase the consequences of an accident

previously evaluated. Additionally, drywell leakage is not the

initiator of any accident evaluated; therefore, changes in the

frequency of the surveillance for drywell leakage does not increase

the probability of any accident evaluated.

Therefore, the proposed changes do not significantly increase

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The requested changes are either administrative changes which

clarify the format of the requirement or change the requirement to

match the design bases of the plant, a change which relocates the

requirement to the Technical Specification Bases, or a change in

surveillance interval. Each of these types of change are discussed

below:

1. The administrative changes in the Technical Specification

requirements do not

[[Page 25706]]

involve a physical alteration of the plant (no new or different type

of equipment will be installed) nor does it change the methods

governing normal plant operation. Thus, this change does not create

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

accident previously evaluated.

2. The proposed relocation of requirements does not involve a

physical alteration of the plant (no new or different type of

equipment will be installed) nor does it change the methods

governing normal plant operation. The proposed change will not

impose or eliminate any requirements. Adequate control of the

information will be maintained in the Technical Specification Bases.

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

or different kind of accident from any accident previously

evaluated.

3. The proposed change modifies the surveillance frequency for

drywell bypass leakage and drywell air lock surveillances. The

changes only impact the test frequency and do not result in any

change in the response of the equipment to an accident. The changes

do not alter equipment design or capabilities. The changes do not

present any new or additional failure mechanisms. The drywell is

passive in nature and the surveillance will continue to verify that

its integrity has not deteriorated. Therefore, the proposed change

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

accident from any accident previously evaluated.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The requested changes are either administrative changes which

clarify the format of the requirement or change the requirement to

match the design bases of the plant, a change which relocates the

requirement to the Technical Specification Bases, or a change in

surveillance interval. Each of these types of changes are discussed

below:

1. The administrative changes in the Technical Specification

requirements do not involve a physical alteration of the plant (no

new or different type of equipment will be installed) nor does it

change the methods governing normal plant operation. Thus, this

change does not cause a significant reduction in the margin of

safety.

2. The relocation of requirements will not reduce a margin of

safety because it has no impact on any safety analysis assumptions.

In addition, the requirements to be transferred from the Technical

Specifications to the Technical Specifications Bases are the same as

the existing Technical Specifications. Since any future changes to

these requirements in the Technical Specifications Bases will be

evaluated per the requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in a margin of safety will be

allowed.

3. The proposed change modifies the surveillance frequency for

drywell bypass leakage and associated air lock surveillances.

Reliability of drywell integrity is evidenced by the measured

leakage rate during past drywell bypass leakage surveillances.

Appropriate design basis assumptions will be upheld, even when

combined with the complementary bypass leakage surveillances as

proposed. Drywell integrity will continue to be tested by means of

the proposed periodic drywell bypass leakage test, performance of

the drywell air lock door latching and interlock mechanism

surveillance, and performance of additional surveillances including

exercising of drywell isolation valves. The combination of these

surveillances will provide adequate assurance that drywell bypass

leakage will not exceed the design basis limit. Margins of safety

would not be reduced unless leakage rates exceeded the design

allowable drywell bypass leakage limit. Therefore, the proposed

change does not cause a significant reduction in the margin of

safety.

Therefore, the proposed changes do not cause 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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120

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

Strawn, 1400 L Street, N.W., 12th Floor, 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: August 11, 1995, as supplemented by

letter dated February 12, 1996.

Description of amendment request: The proposed change will reduce

the minimum reactor coolant cold leg temperature from 544 Degrees F to

541 degrees F in Technical Specification Section 3.2.6, ``Reactor

Coolant Cold Leg Temperature.''

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 involves a 3 deg.F reduction in the minimum

core inlet temperature. This change will not have any impact on the

probability of occurrence of any accident documented in the FSAR.

The impact of this change on the consequences of events

documented in the FSAR has been evaluated. The evaluation

demonstrated that most events are insensitive to the core inlet

temperature. The events that are impacted by lower core inlet

temperature are:

Loss of condenser vacuum (LOCV),

Part length CEA drop,

Single CEA withdrawal within deadband, and

CEA ejection.

The LOCV event has been reanalyzed for the upcoming Cycle (Cycle

8) and the results indicate that the peak RCS pressure remains below

the acceptable limit (110% of the design pressure, i.e., 2750 psia).

The reactivity anomaly events (remaining events) will be reanalyzed

as part of COLSS/CPC setpoint calculations. These calculations will

be performed prior to Cycle 8 startup and will address the impact of

the 3 deg.F reduction on the minimum core inlet temperature. The

CPC/COLSS databases and/or addressable constants will be modified,

as needed due to proposed change, prior to cycle startup.

A qualitative assessment of the impact of the proposed change on

the calculated LOCA blowdown loads that are applied to the major

NSSS components, their supports and the reactor vessel internals was

also performed. This assessment consisted of an evaluation of the

design margins on the major components and a determination of the

impact this lower temperature would have on those margins. The

evaluation concluded that the impact of a 3 deg.F cold leg

temperature reduction will be well within the current design

margins. Therefore, the proposed change will not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

The proposed change to the minimum core inlet temperature does

not involve any change to any equipment or the manner in which the

plant will be operated. Since no hardware modifications or changes

in operation procedures will be made, the proposed change would not

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

any accident previously evaluated. Therefore, the proposed change

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

accident from any accident previously evaluated.

The impact of the proposed change on the Waterford 3 FSAR

analyses have been evaluated. The evaluation showed that the events

that were impacted were important with respect to RCS pressure and

fuel thermal limits. One of the events that was impacted by the

proposed change was the LOCV event. This event was analyzed and the

results showed that the peak RCS pressure remained below the

acceptable limit. The impact of this change on other events

(reactivity anomaly events) will be evaluated as part of the COLSS/

CPC setpoint calculations and the COLSS/CPC databases and/or

addressable constants will modified as needed to account for any

adverse impact on the results of these events due to the proposed

change.

The impact of this change on the Linear Heat Generation Rate

limits which varies as a function of the cold leg temperature, is

accounted for by Technical Specification 3.2.1, ``Linear Heat

Rate''. The impact of this change on LOCA blowdown loads were

evaluated to be insignificant compared to the

[[Page 25707]]

current design margins. Therefore, the proposed change will not

involve a significant reduction in a margin of safety, specifically

fuel thermal limits and RCS pressure limit.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Dates of amendment request: March 20, 1996, and April 23, 1996

Description of amendment request: The licensee proposed to change

the Turkey Point Units 3 and 4 Technical Specifications (TS) to

relocate the requirements for surveillance testing of the water level

and pressure channel instrumentation for the reactor coolant system

accumulators and clarify the remaining TS surveillance tests. These

amendments also modify the existing action statements of TS 3.5.1 for

accumulators to reflect the requirements of NUREG-1431 by requiring a

72-hour period to restore boron concentration if it is not within the

limits, and a 1-hour period to restore any other condition rendering

the accumulators inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed amendments conform to the guidance given in

Enclosure 1 of the NRC GL [Generic Letter] 93-05. The overall

functional capabilities of the Emergency Core Cooling System (ECCS)

accumulators will not be modified by the proposed change. This

amendment will not involve a significant increase in the probability

or consequences of an accident previously evaluated for the

following reasons:

1) The Water Level and Pressure Channel Instrumentation does not

perform a specific safety function, and merely provides an

indicating function. The instrumentation in no way affects the

capability of the accumulators to perform their respective safety

function.

2) The changes in most of the ACTION statements are more

restrictive than current TS requirements due to the one hour vice

four hour completion time, and therefore will not increase the

probability or consequences of a previously evaluated accident. If

one accumulator is inoperable for a reason other than boron

concentration, the accumulator must be returned to OPERABLE status

within 1 hour. In this condition, the required contents of three

accumulators cannot be assumed to reach the core during a Loss Of

Coolant Accident (LOCA). Due to the severity of the consequences

should a LOCA occur in these conditions, the 1 hour completion time

to open the valve, remove power to the valve, or restore the proper

water volume or nitrogen cover pressure ensures that prompt action

will be taken to return the inoperable accumulator to OPERABLE

status. The completion time minimizes the potential for exposure of

the plant to a LOCA under these conditions. The 1 hour requirement

for restoring a closed isolation valve is merely a clarification of

the existing ``immediate'' time requirement.

3) In the case of low-out-of-specification boron concentration

in one accumulator, it must be returned to within the limits within

72 hours. In this condition, ability to maintain subcriticality or

minimum boron precipitation time may be reduced. The boron in the

accumulators contributes to the assumption that the combined ECCS

water in the partially recovered core during the early reflooding

phase of a large break LOCA is sufficient to keep that portion of

the core subcritical. One accumulator below the minimum boron

concentration limit, however, will have no effect on available ECCS

water and an insignificant effect on core subcriticality during

reflood. Boiling of ECCS water in the core during reflood

concentrates boron in the saturated liquid that remains in the core.

In addition, current Turkey Point analysis demonstrate that the

accumulators discharge only a small amount following a large main

steam line break. Therefore, their impact on boron concentration in

the reactor coolant system is minor and not a design limiting event.

Thus, 72 hours is allowed to return the boron concentration to

within limits and does not increase the probability or consequences

of an accident previously evaluated.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the surveillance changes and clarifications, since the proposed

changes do not involve the addition or modification of equipment nor

do they alter the design or operation of affected plant systems.

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

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

system are unchanged by the proposed amendment. The modified

specifications which remove surveillance requirements from the TS to

plant procedures are consistent with the NRC GL 93-05 line-item

improvement guidance do not significantly reduce any of the margins

of safety even though the amount of surveillances is decreased. The

modification of the existing ACTION Statements do not have an

adverse on [sic] affect on the margin of safety for the following

reasons:

1) The SI [Safety Injection] Accumulator Water Level and

Pressure Channel instrumentation performs no safety function.

2) The changes in ACTION statements a) and b) are for the most

part more restrictive than existing TS requirements, the reason

being the removal of instrumentation requirements for operability.

3) In the case of low-out-of-specification boron concentration

in one accumulator, the requirement will be less restrictive, but

the low boron concentration in one accumulator will have no effect

on available ECCS water and an insignificant effect on core

subcriticality during reflood and therefore will not significantly

reduce the margin of safety.

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

this review, it appears that the three standards of 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: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J. Hebdon

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: April 19, 1996

Description of amendment request: The proposed amendment would

include revisions to Technical Specification (TS) 3.3.6.1, ``Primary

Containment and Drywell Isolation Instrumentation; TS

3.3.6.2, ``Secondary Containment Isolation Instrumentation;

TS 3.3.7.1, ``Control Room Ventilation System

Instrumentation; TS 3.6.1.2, ``Primary Containment Air

Locks; TS 3.6.1.3,

[[Page 25708]]

``Primary Containment Isolation Valves; TS 3.6.4.1,

``Secondary Containment; TS 3.6.4.2, ``Secondary Containment

Isolation Dampers; TS 3.6.4.3, ``Standby Gas

Treatment; TS 3.7.3, ``Control Room Ventilation;

and TS 3.7.4, ``Control Room AC System.'' These TSs would be revised to

eliminate CORE ALTERATIONS as an applicable condition for which the

associated Limiting Conditions for Operation (LCO) must be met.

Consistent changes are also proposed for the associated ACTIONS in each

of these LCOs, to reflect the changes in the applicable conditions. The

intent of these proposed changes is to allow certain activities such as

control rod venting, which is considered a CORE ALTERATION in MODE 5,

to be performed without the requirements of the identified LCOs being

met.

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 eliminate CORE ALTERATIONS as an

applicable condition requiring operability of the primary and

secondary containment and control room ventilation system. As stated

in the BASES for the associated Technical Specifications,

operability of these systems is primarily required for mitigation of

the design basis accident - fuel handling accident (DBA-FHA) and

design basis accident - loss of coolant accident (DBA-LOCA). The

performance of CORE ALTERATIONS alone is neither a precursor to, nor

a condition during which these DBAs are postulated to occur. The

proposed changes only delete CORE ALTERATIONS as an applicable

condition for the affected Technical Specifications. All other

applicable MODES or specified conditions, including operations with

the potential for draining the reactor vessels (OPDRVs) and the

movement of irradiated fuel assemblies within the primary or

secondary containment, remain unchanged. Further, the limitations

placed on the handling of light loads are also unchanged. The

Technical Specifications (and the separate requirements imposed on

the handling of light loads) will thus continue to require that

systems or functions designed to mitigate design-basis/previously

evaluated accidents are OPERABLE during the relevant operating MODES

or conditions. On the basis of the above, it is concluded that the

requested amendment will not increase the probability or

consequences of any accident previously evaluated.

2. The proposed changes do not involve any modification to the

plant design or to the operation of plant systems (except to

determine when certain analyzed accident-mitigating systems or

features are required to be OPERABLE). The failure modes considered

for the proposed changes are the same as those previously

considered, therefore, it can be concluded that no new failure modes

will be created. On this basis, the proposed amendment will not

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

any accident previously evaluated.

3. The changes being made to eliminate CORE ALTERATIONS as an

applicable condition for which certain LCOs must be met, do not

eliminate the requirements for operability of those systems or

features assumed to mitigate design-basis or analyzed accidents

during the applicable MODES when such systems or features are

assumed to be available for performing their mitigating function.

The safety margins assumed or established by the accident analyses

for those design-basis events (as described in the accident analyses

of the Clinton Power Station Updated Final Safety Analysis Report)

therefore remain unchanged. Further, the proposed changes do not

impact the controls imposed on the handling of light loads

(including unirradiated fuel assemblies) for ensuring that such

activities cannot result in an event that yields consequences more

severe than those calculated for the DBA-FHA. With respect to

reactivity concerns during refueling operations (MODE 5), all

systems or features required to be OPERABLE for precluding

inadvertent criticality and monitoring reactivity changes will

continue to be required OPERABLE as per the current Technical

Specification requirements. The deletion of CORE ALTERATIONS as an

applicable condition only applies to the noted systems which do not

contribute to precluding reactivity events. Based on the above, the

proposed changes do not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Leah Manning Stetener, 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: May 1, 1996

Description of amendment request: The proposed amendment would

revise the Clinton Power Station (CPS) Operating License and Technical

Specifications (TS) to implement 10 CFR Part 50, Appendix J - Option B,

by referring to Regulatory Guide 1.163, ``Performance-Based Containment

Leak-Test Program.'' Specifically, changes would be made to paragraph

2.D of the Operating License; TS Section 1.1, ``Definitions;'' TS

3.6.1.1, ``Primary Containment;'' TS 3.6.1.1, ``Primary Containment Air

Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs);''

and TS Section 5.5, ``Programs and Manuals.''

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 implements new Option B of 10 CFR 50

Appendix J for performance-based primary containment leakage

testing. The proposed change does not involve a change to the plant

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

affect any parameters or conditions that contribute to the

initiation of any accidents previously evaluated. Thus, the proposed

change cannot increase the probability of any accident previously

evaluated.

The proposed change potentially affects the leak-tight integrity

of the primary containment structure which is designed to mitigate

the consequences of a loss-of-coolant accident (LOCA) by limiting

the release of fission products contained in the post-LOCA primary

containment atmosphere. Functional integrity of the primary

containment must be maintained during and following the peak

transient pressures and temperatures that may result from a LOCA.

Because the proposed change does not alter the plant design,

including the primary containment and primary containment

penetrations, and because it only affects the frequency of measuring

Type A, B, and C leakage without changing the acceptance criteria

for the Type A, B, and C leakage rate tests, the proposed change

does not directly result in an increase in the primary containment

leakage. However, decreasing the test frequency can increase the

probability that an increase in primary containment leakage could go

undetected for an extended period of time. To minimize that

probability, test intervals will be established based on the

performance history of components being tested.

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

provides the technical basis for the NRC's rulemaking to revise

primary containment leakage testing requirements for nuclear power

reactors in 10 CFR 50, Appendix J. NUREG-1493 documents the NRC's

determination that the effect of primary containment leakage on

overall accident risk is minimal since risk is dominated by accident

sequences that result in failure of bypass of primary containment.

NUREG-1493 also documents that increasing the Type A leakage test

intervals would have a minimal impact on public risk, and that Type

B and C tests can identify the vast majority (greater than ninety

five percent) of all leakage paths. Therefore, performance-based

alternatives to current local leakage-testing requirements are

feasible without significant risk impacts.

[[Page 25709]]

Based on the above, IP 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. 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 involves the reduction of

Type A, B, and C test frequency. Except for the method of defining

the test frequency, the methods for performing the actual tests are

not changed. No new accident modes are created by extending the

testing intervals. No safety-related equipment or safety functions

are altered as a result of this change. Thus, extending the test

frequency 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, IP 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 to safety. The proposed change only affects the frequency of

the Type A, B, and C testing. Except for the method of defining the

test frequency, the methods for performing the actual tests are not

changed. However, the proposed change can increase the probability

that an increase in primary containment 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 primary containment is negligible with

the implementation of these proposed changes.

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 the rate of

primary containment leakage. The primary containment isolation

system is designed to limit leakage to La, which is defined by

the CPS Technical Specifications to be 0.65% of primary containment

air weight per day at the calculated peak containment internal

pressure for the design basis loss of coolant accident (Pa).

The limitation on the rate of primary containment leakage is

designed to ensure that 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 primary containment

leakage rate is maintained by continuing to meet the 1.0 La

acceptance criteria. The La value is not being modified by this

proposed change.

Except for the method of defining the test frequency, no change

in the method of testing is being proposed. The Type A, B, and C

tests will continue to be done at full pressure (Pa) or

greater. Other programs are in place to ensure that proper

maintenance and repairs are performed during the service life of the

primary containment and systems and components penetrating the

primary containment.

As a result, IP has concluded that the proposed change will not

result in a significant reduction in a margin of safety.

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

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

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

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 Stetener, Vice President,

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

Decatur, Illinois 62525

NRC Project Director: Gail H. Marcus

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: January 25, 1996

Description of amendment request: The amendment proposes to extend

instrumentation and miscellaneous surveillance test intervals (STI) to

support 24-month operating cycles. Additionally, this application

proposes: (1) to revise the Trip Level Settings for Emergency Bus Loss

of Voltage and Degraded Voltage Instrumentation, (2) to revise the

Reactor Protection System (RPS) Normal Supply Electrical Protection

Assembly (EPA) Undervoltage Trip Setpoint, and (3) to make editorial

revisions, clarification and Bases changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed STI changes evaluated in Section IV.A do not

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

these systems function, and will not degrade the performance of the

plant safety systems. Proposed instrument setpoint changes ensure

that plant safety limits are not exceeded due to instrument drift

predicted for the longer calibration interval. The type of testing

and the corrective actions required if the subject surveillances

fail remains the same. The proposed changes do not adversely affect

the reliability of these systems or affect the ability of the

systems to meet their design objectives. A historical review of

surveillance test results supports these conclusions.

The Trip Level Setpoint changes evaluated in Section IV.B ensure

that the related systems perform as assumed in the transient and

accident analysis by ensuring that plant safety limits are not

exceeded due to instrument drift predicted for the longer

calibration interval. The changes do not alter the system function,

and will not degrade the performance of plant safety systems. The

proposed Trip Level Setting changes do not adversely affect the

reliability of these systems or adversely affect the ability of

these systems to meet their design objectives.

The editorial, clarification and Bases changes evaluated in

Section IV.C propose enhancements that clarify the Technical

Specifications requirements and are editorial in nature. These

changes do not alter any Technical Specification requirement, do not

involve physical changes to the plant, or alter any operational

setpoints. There are no safety implications in these proposed

changes.

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

from any accident previously evaluated.

The proposed STI changes evaluated in Section IV.A do not modify

the design or operation of the plant, therefore, no new failure

modes are introduced. Proposed instrument setpoint changes ensure

that plant safety limits are not exceeded due to instrument drift

resulting from the longer calibration interval. No changes are

proposed to the type and method of testing performed, only to the

length of the surveillance test interval. Past equipment performance

and on-line testing indicate that longer test intervals will not

degrade these systems. A historical review of surveillance test

results supports these conclusions.

The Trip Level Setpoint changes evaluated in Section IV.B ensure

that the related systems perform as assumed in the transient and

accident analysis by ensuring that plant safety limits are not

exceeded due to instrument drift predicted for the longer

calibration interval. The changes do not alter the system function,

introduce any new failure modes, and will not degrade the

performance of plant safety systems. The proposed Trip Level Setting

changes do not adversely affect the reliability of these systems or

adversely affect the ability of these systems to meet their design

objectives.

The editorial, clarification and Bases changes evaluated in

Section IV.C propose enhancements that clarify the Technical

Specifications requirements and are editorial in nature. These

changes do not alter any Technical Specification requirement, do not

involve physical changes to the plant, or alter any operational

setpoints. There are no safety implications in these proposed

changes.

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

Although the proposed STI changes evaluated in Section IV.A will

result in an increase in the interval between surveillance tests,

the impact on system reliability is minimal. This is based on more

frequent on-line testing and the redundant design of the evaluated

systems. A review of past surveillance history has shown no evidence

[[Page 25710]]

of failures which would significantly impact the reliability of

these systems. Operation of the plant remains unchanged by these

proposed STI extensions. The assumptions in the Plant Licensing

Basis are not adversely impacted. Therefore, the proposed changes do

not result in a significant reduction in the margin of safety.

The Trip Level Setpoint changes evaluated in Section IV.B ensure

that the related systems perform as assumed in the transient and

accident analysis by ensuring that plant safety limits are not

exceeded due to instrument drift predicted for the longer

calibration interval. The changes do not alter the system function,

introduce any new failure modes, and will not degrade the

performance of plant safety systems. The proposed Trip Level Setting

changes do not adversely affect the reliability of these systems or

adversely affect the ability of these systems to meet their design

objectives.

The editorial, clarification and Bases changes evaluated in

Section IV.C propose enhancements that clarify the Technical

Specifications requirements and are editorial in nature. These

changes do not alter any Technical Specification requirement, do not

involve physical changes to the plant, or alter any operational

setpoints. There are no safety implications in these proposed

changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Susan Frant Shankman, Acting

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: April 24, 1996

Description of amendment request: This amendment proposes to

relocate Technical Specification (TS) 3.11.B/4.11.B ``Crescent Area

Ventilation'' and associated Bases from the TS to an Authority

controlled procedure.

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:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment will not involve a significant hazards

consideration as defined in 10 CFR 50.92, based on the following:

(1) These changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated

because:

No modifications, no changes to operating procedure

requirements, and no reduction in equipment reliability are being

made as a result of these changes. Operating limitations will

continue to be imposed, and required surveillance will continue to

be performed in accordance with regulations, and written procedures

and instructions that are auditable by the [Nuclear Regulatory

Commission] NRC. Crescent Area Ventilation operability and testing

requirements will continue to be an integral part of FitzPatrick

plant operation.

Although future changes to the Crescent Area Ventilation system

will no longer be controlled by 10 CFR 50.90, proposed changes will

be evaluated under 10 CFR 50.59 and plant procedures. Programmatic

controls will continue to assure that Crescent Area Ventilation

system changes will not adversely affect [Emergency Core Cooling

System] ECCS or [Reactor Core Isolation Cooling] RCIC system

operability. As such, there is no significant increase in the

probability or consequences of an accident previously evaluated.

(2) These changes do not create the possibility of a new or

different type of accident previously evaluated because:

No modifications, no changes to operating procedure

requirements, and no reduction in equipment reliability are being

made as a result of these changes. Compliance with Crescent Area

Ventilation system operability and surveillance requirements will be

assured by maintaining them in an Authority controlled procedure.

Changes to the Crescent Area Ventilation system will be subject to

the requirements of 10 CFR 50.59. Therefore, the proposed changes do

not introduce any failure mechanism of a different type than those

previously evaluated since there are no changes being made to the

facility and do not create the possibility of a new or different

type of accident previously evaluated.

(3) The proposed amendment does not involve a reduction in a

margin of safety because:

The Crescent Area Ventilation system supports Core Spray, [Low

Pressure Coolant Injection] LPCI mode of [Residual Heat Removal]

RHR, containment cooling mode of RHR, [High Pressure Coolant

Injection] HPCI, and RCIC operability, and Crescent Area Ventilation

system inoperability does affect these systems. As a result, the

requirement for Crescent Area Ventilation to be operable for these

systems to be considered operable is implicit in TS Sections 3.5.A,

3.5.B, 3.5.C, 3.5.E, and the definition of OPERABLE contained in TS

Section 1.0.J. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Susan Frant Shankman, Acting

Public Service Electric & Gas Company, Docket No. 50-311, Salem

Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

Date of amendment request: May 7, 1996

Description of amendment request: The proposed amendment involves a

one-time change to Technical Specification (TS) 3/4.7.6, ``Control Room

Emergency Air Conditioning System.'' The change would permit refueling

of Salem, Unit 2, with the Control Room Emergency Air Conditioning

System (CREACS) inoperable in Modes 5 and 6. The change will expire

after the completion of the Control Room and CREACS upgrade, which is

currently in progress, and the restart and entry into Mode 4 of Unit 2

from the current outage.

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 CREACS is not an accident initiator. CREACS functions post-

accident to provide cooling for Control Room equipment and

habitability for operations personnel. Therefore, CREACS has no

influence on the probability of any of the previously evaluated

accidents or the other events evaluated as listed below.

Event

Fuel Handling Accident (Salem)

Waste Gas or Volume Control Tank Failures

Uncontrolled Boron Dilution

Loss of Offsite Power

Fuel Handling Accident (Hope Creek)

Liquid and Gaseous Waste Releases (Hope Creek)

Loss of Coolant Accident (LOCA) (Hope Creek)

Chemical Storage

Barge Collision

Control Room Internal and External Fire

Loss of Spent Fuel Pool Cooling

Loss of Decay Heat Removal

The Control Area Air Conditioning System (CAACS) and other

measures will be

[[Page 25711]]

available to maintain Control Room Envelope (CRE) ambient

temperatures and habitability.

The proposed one-time change does not impact the consequences of

an accident previously evaluated based on the following discussions.

The fuel has decayed to such low levels for more than six months

that doses associated with the fuel handling accident are well

within the limits of GDC [General Design Criteria] 19. There is

insufficient activity remaining in either gaseous waste storage or

liquid waste storage to force a Control Room evacuation. In the

event of a Loss of Offsite Power (LOOP), uncontrolled boron dilution

event, loss of spent fuel pool cooling or loss of decay heat

removal, CREACS is not required in Modes 5 or 6 to mitigate the

consequences of this event and CRE habitability will be maintained.

For a Hope Creek fuel handling accident, gaseous radwaste

release of LOCA, dose to Salem Control Room personnel will not

exceed GDC 19 limits. PSE&G [Public Service Electric & Gas] will

maintain the CAACS [Control Area Air Conditioning System] outside

air intakes either isolated or capable of being isolated in the

event of a Hope Creek LOCA. The Hope Creek Event Classification

Guide (ECG) requires notification of the Salem Control Room in the

event of an emergency that has the potential to result in a

radioactive release. The Salem Control Room will isolate the outside

air intakes if isolation has not already been accomplished.

For the other events evaluated, the need for evacuation is not

considered credible for any event with the exception of an internal

or external fire. However, the possibility of evacuation of the CRE

in the event of an internal or external fire would be no different

whether or not CREACS is operating. In the event of an internal

fire, CAACS will remain in operation to provide purging of the CRE.

For the case of a possible external fire, the need for evacuation is

not considered credible because of the short duration of the CREACS

outage and improbability of the factors which are necessary to

require an evacuation of the Control Room (i.e. wind direction, wind

speed, amount of smoke). If an external fire is detected, operator

action will be taken to isolate the CRE from outside air while CAACS

remains available. In the unlikely event that the Control Room would

become uninhabitable due to smoke in the atmosphere, evacuation

procedures would be followed as in the case of the internal fire.

The one chemical storage type event which might impact the

Control Room, rupture of an ammonium hydroxide tanker, is precluded

by administrative controls such that no ammonium hydroxide tanker

deliveries will be allowed during the system upgrade period.

The CAACS will maintain the current design function and TS Bases

requirements of the CREACS that the ambient air temperature does not

exceed the allowable temperature for continuous duty rating for

equipment and instrumentation cooled by the system for the combined

CRE. The CAACS will be maintained functional while modification to

the CREACS is ongoing to provide cooling during normal operation and

under postulated accident conditions. Should the temperature in the

CRE exceed allowable levels (85 Degrees F), administrative controls

will be in place to require restoration of the temperature to within

acceptable levels using CAACS, and prevent any Core Alteration

activities or positive reactivity changes until the temperature is

restored to acceptable levels.

Therefore, the proposed one-time TS change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The CREACS is not an accident initiator. CREACS functions post-

accident to provide cooling for Control Room equipment and

habitability for operations personnel. Therefore, CREACS

inoperability during Modes 5 and 6 will not result in the creation

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

evaluated. All pertinent accidents have been assessed and no other

scenarios dealing with fuel movement, or the need for an operable

CREACS in Mode 5 or 6, have been deemed credible.

Therefore, the proposed one-time 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.

The proposed one-time change does not significantly reduce the

margin of safety as defined in the Bases for the TS because (1)

there is no credible event as analyzed in Salem UFSAR [updated final

safety analysis report] Chapter 15 which can cause an unacceptable

environment in the CRE since the fuel has been decaying for at least

six months, (2) fuel movement inside the Fuel Handling Building

(FHB) is restricted in accordance with plant TS unless FHB

ventilation is operable, (3) dose to Salem control room personnel

from a potential Hope Creek fuel handling accident, gaseous radwaste

release or Loss of Coolant Accident will not exceed GDC 19 limits

(4) the one event which might impact the Control Room, rupture of an

ammonium hydroxide tanker, is precluded by administrative controls

such that no ammonium hydroxide tanker deliveries will be allowed

during the CREACS upgrade period, and (5) in the unlikely event that

Control Room evacuation is required, there is no impact on operator

ability to mitigate the consequences of an accident in the current

plant configuration.

Therefore, the proposed one-time TS 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph

M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: March 29, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification 3/4.4.6 ``Steam Generators'' and its

associated Bases. Specifically, the steam generator repair limit would

be modified to clarify that the appropriate method for determining

serviceability for tubes with outside diameter stress corrosion

cracking at the tube support plate is by a methodology that more

reliably assesses structural integrity. This amendment request is in

accordance with NRC's Generic Letter 95-05, ``Voltage-Based Repair

Criteria for 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:

Operation of Farley units 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 standing tubes at

room temperature conditions shows burst pressures as high as

approximately 5000 psi for indications of outer diameter stress

corrosion cracking with voltage measurements as high as 26.5 volts.

Burst testing performed on pulled tubes, including tubes pulled from

Farley Unit 2, with up to 7.5 volt indications show burst pressures

in excess of 5300 psi at room temperature. As stated earlier, tube

burst criteria are inherently satisfied during normal operating

conditions by the presence of the tube support plate. Furthermore,

correcting for the effects of temperature on material properties and

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

temperature), tube burst capability significantly exceeds the R.G.

[Regulatory Guide] 1.121 criterion requiring the maintenance of a

margin of 1.43 times the steam line break pressure differential on

tube burst if through-wall cracks are present without regard to the

presence of the tube support plate. Considering the existing data

base, this criterion is satisfied with bobbin coil indications with

signal amplitudes over twice the 2.0 volt voltage-based repair

criteria, regardless of the indicated depth measurement. This

structural limit is based on a lower 95% confidence level limit of

the

[[Page 25712]]

data at operating temperatures. The 2.0 volt criterion provides a

conservative margin of safety to the structural limit considering

expected growth rates of outside diameter stress corrosion cracking

at Farley. Alternate crack morphologies can correspond to a voltage

so that a unique crack length is not defined by a burst pressure to

voltage correlation. However, relative to expected leakage during

normal operating conditions, no field leakage has been reported from

tubes with indications with a voltage level of under 7.7 volts for a

3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as

compared to the 2.0 volt proposed voltage-based tube repair limit).

Thus, the proposed amendment does not involve a significant increase

in the probability or consequences of an accident.

Relative to the expected leakage during accident condition

loadings, the accidents that are affected by primary-to-secondary

leakage and steam release to the environment are Loss of External

Electrical Load and/or Turbine Trip, Loss of All AC Power to Station

Auxiliaries, Major Secondary System Pipe Failure, Steam Generator

Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a

Control Rod Drive Mechanism Housing. Of these, the Major Secondary

System Pipe Failure is the most limiting for Farley in considering

the potential for off-site doses. The offsite dose analyses for the

other events which model primary-to secondary leakage and steam

releases from the secondary side to the environment assume that the

secondary side remains intact. The steam generator tubes are not

subjected to a sustained increase in differential pressure, as is

the case following a steam line break event. This increase in

differential pressure is responsible for the postulated increase in

leakage and associated offsite doses following a steam line break

event. In addition, the steam line break event results in a bypass

of containment for steam generator leakage. Upon implementation of

the voltage-based repair criteria, it must be verified that the

expected distributions of cracking indications at the tube support

plate intersections are such that primary-to-secondary leakage would

result in site boundary dose within the current licensing basis.

Data indicate that a threshold voltage of 2.8 volts could result in

through-wall cracks long enough to leak at steam line break

conditions. Application of the proposed repair criteria requires

that the current distribution of a number of indications versus

voltage be obtained during the refueling outages. The current

voltage is then combined with the rate of change in voltage

measurement and a voltage measurement uncertainty to establish an

end of cycle voltage distribution and, thus, leak rate during steam

line break pressure differential. The leak rate during a steam line

break is further increased by a factor related to the probability of

detection of the flaws. If it is found that the potential steam line

break leakage for degraded intersections planned to be left in

service coupled with the reduced allowable specific activity levels

result in radiological consequences outside the current licensing

basis, then additional tubes will be plugged or repaired to reduce

steam line break leakage potential to within the acceptance limit.

Thus, the consequences of the most limiting design basis accident

are constrained to present licensing basis limits.

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 voltage-based tube support plate

elevation steam generator tube repair criteria does not introduce

any significant changes to the plant design basis. Use of the

criteria does not provide a mechanism that could result in an

accident outside of the region of the tube support plate elevations.

Neither a single or multiple tube rupture event would be expected in

a steam generator in which the repair criteria have been applied

during all plant conditions. The bobbin probe signal amplitude

repair criteria are established such that operational leakage or

excessive leakage during a postulated steam line break condition is

not anticipated. Southern Nuclear has previously implemented a

maximum leakage limit of 150 gpd per steam generator. The R.G. 1.121

criterion for establishing operational leakage limits that require

plant shutdown are based upon leak-before-break considerations to

detect a free span crack before potential tube rupture. 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. R.G.

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 longest permissible crack is the

length that provides a factor of safety of 1.43 against bursting at

steam line break pressure differential. A voltage amplitude of

approximately 9 volts for typical outside diameter stress corrosion

cracking corresponds to meeting this tube burst requirement at the

95% prediction interval on the burst correlation. Alternate crack

morphologies can correspond to a voltage so that a unique crack

length is not defined by the burst pressure versus voltage

correlation. Consequently, a typical burst pressure versus through-

wall crack length correlation is used below to define the ``longest

permissible crack'' for evaluating operating leakage limits.

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

at 1.43 times steam line break pressure differential and steam line

break conditions are about 0.54 inch and 0.84 inch, respectively.

Normal leakage for these crack lengths would range from about 0.4

gallons per minute to 4.5 gallons per minute, respectively, while

lower 95% confidence level leak rates would range from about 0.06

gallons per minute to 0.6 gallons per minute, respectively.

An operating leak rate of 150 gpd per steam generator has been

implemented. This leakage limit provides for detection of 0.4 inch

long cracks at nominal leak rates and 0.6 inch long cracks at the

lower 95% confidence level leak rates. Thus, the 150 gpd limit

provides for plant shutdown prior to reaching critical crack lengths

for steam line break conditions at leak rates less than a lower 95%

confidence level and for three times normal operating pressure

differential at less than nominal leak rates.

Considering the above, the implementation of voltage-based

plugging criteria will not create the possibility of a new or

different kind of accident from any previously evaluated.

3) The proposed license amendment does not involve a significant

reduction in margin of safety.

The use of the voltage-based tube support plate elevation repair

criteria is demonstrated to maintain steam generator tube integrity

commensurate with the requirements of Generic Letter 95-05 and R.G.

1.121. R.G. 1.121 describes a method acceptable to the NRC staff for

meeting GDC [Generic Design Criteria] 2, 14, 15, 31, and 32 by

reducing the probability of 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 removed from service. Upon implementation of the criteria,

even under the worst case conditions, the occurrence of outside

diameter stress corrosion cracking 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 most limiting

effect would be a possible increase in leakage during a steam line

break event. Excessive leakage during a steam line break event,

however, is precluded by verifying that, once the criteria are

applied, the expected end of cycle distribution of crack indications

at the tube support plate elevations would result in minimal, and

acceptable primary to secondary leakage during the event and, hence,

help to demonstrate radiological conditions are less than an

appropriate fraction of the 10 CFR [Part] 100 guideline.

The margin to burst for the tubes using the voltage-based repair

criteria is comparable to that currently provided by existing

Technical Specifications.

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

accident] + SSE [safe-shutdown earthquake] 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 either the LOCA rarefaction wave and/or 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 [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 (PCT). Second, there is a potential the

partial through-wall cracks in tubes could progress to through-wall

cracks during tube deformation or collapse or that short through-

[[Page 25713]]

wall indications would leak at significantly higher leak rates than

included in the leak rate assessments.

Consequently, a detailed leak-before-break analysis was

performed and it was concluded that the leak-before-break

methodology (as permitted by GDC 4) is applicable to the Farley

reactor coolant system primary loops and, thus, the probability of

breaks in the primary loop piping is sufficiently low that they need

not be considered in the structural design basis of the plant.

Excluding breaks in the RCS primary loops, the LOCA loads from the

large branch line breaks were analyzed at Farley and were found to

be of insufficient magnitude to result in steam generator tube

collapse or significant deformation.

Regardless of whether or not leak-before-break is applied to the

primary loop piping at Farley, any flow area reduction is expected

to be minimal (much less than 1%) and PCT margin is available to

account for this potential effect. Based on analyses' results, no

tubes near wedge locations are expected to collapse or deform to the

degree that secondary to primary in-leakage would be increased over

current expected levels. For all other steam generator tubes, the

possibility of secondary-to-primary leakage in the event of a LOCA +

SSE event is not significant. In actuality, the amount of secondary-

to-primary leakage in the event of a LOCA + SSE is expected to be

less than that originally allowed, i.e., 500 gpd per steam

generator. Furthermore, secondary-to-primary in-leakage would be

less than primary-to-secondary leakage for the same pressure

differential since the cracks would tend to tighten under a

secondary-to-primary pressure differential. Also, the presence of

the tube support plate is expected to reduce the amount of in-

leakage.

Addressing the R.G. 1.83 considerations, implementation of the

tube repair criteria is supplemented by 100% inspection requirements

at the tube support plate elevations having outside diameter stress

corrosion cracking indications, reduced operating leakage limits,

eddy current inspection guidelines to provide consistency in voltage

normalization, and rotating probe inspection requirements for the

larger indications left in service to characterize the principle

degradation mechanism as outside diameter stress corrosion cracking.

As noted previously, implementation of the tube support plate

elevation repair criteria will decrease the number of tubes that

must be taken out of service with tube plugs or repaired. The

installation of steam generator tube plugs or tube sleeves would

reduce the RCS flow margin, thus implementation of the voltage-based

repair criteria will maintain the margin of flow that would

otherwise be reduced through increased tube plugging or sleeving.

Considering the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the Final Safety Analysis Report 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: 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 No. 50-364, Joseph

M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: April 22, 1996

Description of amendment request: The proposed amendment would

implement a new F* criterion based on maintaining existing safety

margins for steam generator tube structural integrity concurrent with

allowance for NDE (nondestructive examination) eddy current

uncertainty.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change retains the existing margin in the F*

distance used to meet regulatory guidance of draft Regulatory Guide

1.121 and only changes the amount of assumed NDE eddy current

uncertainty based on the type of eddy current technology utilized in

the inspection. Therefore, there is no 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 previously evaluated. WCAP

11306, Revision 2, ``Tubesheet Region Plugging Criterion for the

Alabama Power Company Farley Nuclear Station Unit 2 Steam

Generators,'' provides adequate basis for the F* distance proposed

of 1.54 plus allowance for eddy current uncertainty measurement.

Since the value of 1.54 inches was used in the analysis no new or

different kind of accident from any accident previously evaluated

will be created.

3. The proposed change does not involve a significant reduction

in a margin safety. Since the value of 1.54 inches already is used

in the steam generator tube pull out analysis, there is no

significant change to a margin 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: 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

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, MissouriDate of application request: February 23,

1996, as supplemented by letter dated April 24, 1996.

Description of amendment request: The amendment would add a

footnote in the license for Callaway Plant, Unit No. 1 to indicate that

Union Electric Company has entered into a merger agreement with CIPSCO

Incorporated which provides for Union Electric Company to become a

wholly-owned operating company of Ameren Corporation, a registered

public utility holding company under the Public Utility Holding Company

Act of 1935, as amended. After the merger, Union Electric Company would

continue to own and operate the Callaway Plant as an operating company

subsidiary of Ameren Corporation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change does not affect accident initiators or

assumptions. The radiological consequences of any accident

previously evaluated remain unchanged. The change is an

administrative change to reflect Union Electric's status as an

operating company subsidiary of Ameren.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not reduce the margin of safety assumed

in any accident analysis or affect any safety limits. The change is

administrative and reflects Union Electric's status as an operating

company subsidiary of Ameren.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not reduce the margin of safety assumed

in any accident

[[Page 25714]]

analysis or affect any safety limits. The change is administrative

and reflects Union Electric's status as an operating company

subsidiary of Ameren.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

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

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: William H. Bateman

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 30, 1996

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

3.1.b.1, its associated bases, and Figure TS 3.1-4 by extending the low

temperature overpressure protection (LTOP) requirements through the end

of operating cycle 33 or 33.41 effective full power years. The only

technical change being proposed is the substitution of end of life

fluence for the end of operating cycle 21 fluence.

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 was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The LTOP setpoint and revised P/T [pressure/temperature] limits

reflected in proposed Figure TS 3.1-4 ensure that the Appendix G

pressure/temperature limits are not exceeded, and therefore, help

ensure that RCS integrity is maintained. The changes do not modify

the reactor coolant system pressure boundary, nor make any physical

changes to the facility design, material, construction standards, or

setpoints. The LTOP valve setpoint remains set at 500 psi. The LTOP

enabling temperature based on Figure TS 3.1-2 is 338 deg.F and is

more conservative than a value of 303 deg. Figure TS 3.1-4. The LTOP

enabling temperature based on Figure TS 3.1-2 remains unchanged by

this PA [proposed amendment]. The probability of a LTOP event

occurring is independent of the pressure-temperature limits for the

RCS pressure boundary. Therefore, the probability of a LTOP event

occurring remains unchanged.

The calculation of pressure temperature limits in accordance

with approved regulatory methods provides assurance that reactor

pressure vessel fracture toughness requirements are met and the

integrity of the RCS [reactor coolant system] pressure boundary is

maintained. Similar methodology was used in calculations to support

approved amendment 120 to the Kewaunee Technical Specifications

dated April 26, 1995. The material property basis, including

chemistry factor and initial reference temperature for the

unirradiated material (RTNDT), used for this PA is the same as

that used in the current TS. The only technical change being made in

this PA is the use of end of life fluence.

The use of predicted fluence values through the end of operating

cycle 33 is appropriately considered within the calculations in

accordance with standard industry methodology previously docketed

under WCAP 13227 and WCAP 14279. The neutron exposure projections

utilized for calculation of the reference temperature were

multiplied by a factor of 1.11 to adjust for biases observed between

cycle specific calculations and the results of neutron dosimetry for

the four surveillance capsules removed from the KNPP reactor. The

factor of 1.11 was derived by taking the average of the measured to

calculation (M/C) flux ratios obtained from the dosimetry results of

capsules V, R, P, and S removed from the KNPP reactor vessel. The

resulting effect of using predicted fluence values through the end

of cycle 33 instead of cycle 21 is to require the plant to evaluate

LTOP transients to more limiting requirements. The proposed PT

limits are shifted to a lower pressure and higher temperature, which

is more conservative.

The changes do not adversely affect the integrity of the RCS

such that its function in the control of radiological consequences

is affected. In addition, the changes do not affect any fission

barrier. The changes do not degrade or prevent the response of the

LTOP relief valve or other safety related system to accidents

described in Chapter 14 of the USAR. In addition, the changes do not

alter any assumption previously made in the radiological

consequences evaluations nor affect the mitigation of the

radiological consequences of an accident described in the USAR.

Therefore, the consequences of an accident previously evaluated in

the USAR will not be increased.

Thus, the operation of KNPP Unit 1 in accordance with the PA

does not involve a significant increase in the probability or

consequences of any accident previously evaluated.

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

from an accident previously evaluated.

The Appendix G pressure temperature limitations were prepared

using methods derived from the ASME Boiler and Pressure Vessel Code

and the criteria set forth in NRC Regulatory Standard Review Plan

5.3.2. The changes do not cause the initiation of any accident nor

create any new credible limiting failure for safety-related systems

and components. The changes do not result in any event previously

deemed incredible being made credible. As such, it does not create

the possibility of an accident different than any evaluated in the

USAR.

The changes do not have any effect on the ability of the safety-

related systems to perform their intended safety functions. The

changes do not create failure modes that could adversely impact

safety-related equipment. Therefore, it will not create the

possibility of a malfunction of equipment important to safety

different than previously evaluated in the USAR. Thus, the PA does

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

from any accident previously evaluated.

The use of Paragraph (c)(2)(ii)(A) of 10 CFR 50.61, initial

reference temperature of -50 deg.F, and the fluence values through

EOC [end of cycle] 33 does not modify the reactor coolant system

pressure boundary, nor make any physical changes to the LTOP

setpoint or system design. Proposed Figure TS 3.1-4 was prepared in

accordance with regulatory requirements and requires evaluation of

LTOP events to more limiting requirements of neutron exposure

projections of 33.41 EFPY instead of 18.40 EFPY.

Therefore, the PA does not create the possibility of a new or

different type of accident from any accident previously evaluated.

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

The Appendix G pressure temperature limitations were prepared

using methods derived from the ASME Boiler and Pressure Vessel Code

and the criteria set forth in NRC Regulatory Standard Review Plan

5.3.2. These documents along with the calculational limitations

specified in 10 CFR 50.61 are an acceptable method for implementing

the requirements of 10 CFR 50 Appendices G and H. Inherent

conservatism in the P/T limits resulting from these documents

include:

a. An assumed defect in the reactor vessel wall with a depth

equal to 1/4 of the thickness of the vessel wall (1/4T) and a length

equal to 1-1/2 times the thickness of the vessel wall.

b. Assumed reference flaw oriented in b

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