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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from April 24, 1999, through May 7, 1999. The

last biweekly notice was published on May 5, 1999.

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 and

Directives Branch, Division of Administration 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 18, 1999, 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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

Station, Plymouth County, Massachusetts

Date of amendment request: March 3, 1999.

Description of amendment request: The proposed amendment would

change the reactor vessel (RV) surveillance capsule pull interval from

approximately 15 effective full power (EFPY) years to 18 EFPY in

Technical Specification (TS) Table 4.6-3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below: The operation of Pilgrim in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

Pilgrim plant's physical configuration and operational practices are

not changed by this proposed change. The licensee is only proposing to

change the TS withdrawal schedule for the RV surveillance capsule. This

change does not affect any of the current accident mitigation features

of the facility or the sequence of any accidents previously analyzed.

For the reasons given above, deferral of withdrawal of Pilgrim's second

capsule for at least one additional cycle (or 3 EFPY) does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

The operation of Pilgrim in accordance with the proposed amendment

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

from any accident previously evaluated. As discussed in the above

narrative, the deferral of the second capsule pull at Pilgrim does not

change any of the design features or operation of the facility but does

defer a TS surveillance. Pilgrim's current TS pressure-temperature (P-

T) curves are conservative and will remain so even if the RV

surveillance capsule is not pulled this outage. The data from the first

RV capsule supports this conclusion. Because the RV capsule pull

schedule is being deferred, the P-T curves, which can be modified based

on the data from the RV capsule surveillance, will not be changed. The

deferral of the withdrawal of Pilgrim's second RV surveillance capsule

does not change the design features or operation of the facility and

the existing P-T curves have not changed, therefore, the TS change will

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

any accident previously evaluated.

The operation of Pilgrim in accordance with the proposed amendment

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

The capsule pull is a surveillance technique that provides data for

modification of the P-T curves. The methods used to develop the

temperatures associated with these curves are regarded as conservative.

The data from the first RV capsule supported this conclusion. Because

the P-T curves have not changed and have been determined to be

conservative, the margins of safety that were previously established

have not changed. Therefore, deferral of the withdrawal of Pilgrim's

second RV surveillance capsule will not involve a significant reduction

in the margin of safety.

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

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

that the

[[Page 27317]]

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 132

South Street, Plymouth, Massachusetts 02360.

Attorney for licensee: J. Fulton, Boston Edison Company, 800

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

NRC Section Chief: James W. Clifford.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: March 30, 1999.

Description of amendment request: The proposed amendment would

revise Section 4.0, Surveillance Requirements, of the Technical

Specifications (TSs). Specifically, Section 4.0.2 would be added to

allow a 24-hour grace period for performing inadvertently missed

surveillance.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response: No. This proposed change will result in either the

plant condition either remaining unchanged (i.e., the system or

component is declared operable) or in the plant proceeding to a

shutdown condition (i.e., the system or component is declared

operable). If at the end of the 24-hour interval, it is necessary to

proceed to shutdown, this shutdown is indistinguishable from any

shutdown where a system or component is declared inoperable.

Allowing an additional 24 hours to perform the surveillance balances

the risks associated with an allowance for completing the

surveillance within this 24-hour period against the risks associated

with the potential for a plant upset and challenge to safety systems

when the alternative is a shutdown to comply with the action

requirements before the surveillance can be completed. Therefore,

the proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated?

Response: No. This proposed change will result in either the

plant condition either remaining unchanged (i.e., the system or

component is declared operable) or in the plant proceeding to a

shutdown condition (i.e., the system or component is declared

operable). If at the end of the 24-hour interval, it is necessary to

proceed to shutdown, this shutdown is indistinguishable from any

shutdown where a system or component is declared inoperable.

Allowing an additional 24 hours to perform the surveillance balances

the risks associated with an allowance for completing the

surveillance within this 24-hour period against the risks associated

with the potential for a plant upset and challenge to safety systems

when the alternative is a shutdown to comply with the action

requirements before the surveillance can be completed. Therefore,

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

different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: No. This proposed change will result in either the

plant condition either remaining unchanged (i.e., the system or

component is declared operable) or in the plant proceeding to a

shutdown condition (i.e., the system or component is declared

operable). If at the end of the 24-hour interval, it is necessary to

proceed to shutdown, this shutdown is indistinguishable from any

shutdown where a system or component is declared inoperable.

Allowing an additional 24 hours to perform the surveillance within

this 24-hour period against the risks associated with the potential

for a plant upset and challenge to safety systems when the

alternative is a shutdown to comply with the action requirements

before the surveillance can be completed . Therefore, the proposed

change does not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Section Chief: S. Singh Bajwa.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of amendment request: July 22 and October 22, 1998; May 6,

1999.

Description of amendment request: The amendments would revise the

Technical Specifications (TS) to reflect the licensee's planned use of

fuel supplied by Westinghouse. The staff has published a Notice of

Consideration of Issuance of Amendments and Proposed No Significant

Hazards Consideration Determination on November 18, 1998 (63 FR 64108)

covering the July 22 and October 22, 1998, submittals. In the May 6,

1999, submittal the licensee proposed to expand the original amendment

request, revising Section 5.6.5 of the Technical Specifications.

Section 5.6.5 specifies a list of NRC-approved topical reports that the

licensee is required to use to determine reactor core operating limits.

The licensee proposed to update this list to show the current approval

status of these topical reports.

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 for the proposed changes conveyed by the May 6, 1999,

submittal. The NRC staff has reviewed the licensee's analyses against

the standards of 10 CFR 50.92(c). The NRC-staff's analysis is presented

below.

First Standard

No. The proposed changes to Section 5.6.5 will not affect the

safety function and will not involve any change to the design or

operation of any plant system or component. The topical reports were

previously approved by the NRC staff under separate licensing actions.

The use of methodologies in these approved topical reports will ensure

that previously evaluated accidents remain bounding. Therefore, no

accident probabilities or consequences will be impacted.

Second Standard

No. The proposed changes would not lead to any hardware or

operating procedure change. Hence, no new equipment failure modes or

accidents from those previously evaluated will be created.

Third Standard

No. Margin of safety is associated with confidence in the design

and operation of the plant; specifically, the ability of the fission

product barriers to perform their design functions during and following

an accident. The proposed changes to Section 5.6.5 do not involve any

change to plant design, operation, or analysis. Thus, the margin of

safety previously analyzed and evaluated is maintained.

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

CFR 50.92(c) are satisfied for the proposed changes to Section 5.6.5.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

[[Page 27318]]

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: April 5, 1999.

Description of amendment request: The proposed amendments would

provide revised spent fuel pool storage configurations, revised spent

fuel pool storage criteria, and revised fuel enrichment and burnup

requirements which take credit for soluble boron in maintaining

acceptable margins of subcriticality in the spent fuel storage pools.

Also, the proposed amendments would provide additional criteria for

ensuring acceptable levels of subcriticality in the spent fuel storage

pools.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will the change involve a significant increase in the

probability or consequence of an accident previously evaluated?

No, based upon the following:

Dropped Fuel Assembly

There is no significant increase in the probability of a fuel

assembly drop accident in the spent fuel pools when considering the

degradation of the Boraflex panels in the spent fuel pool racks

coupled with the presence of soluble boron in the spent fuel pool

water for criticality control. The handling of the fuel assemblies

in the spent fuel pool has always been performed in borated water,

and the quantity of Boraflex remaining in the racks has no affect on

the probability of such a drop accident.

The criticality analysis showed that the consequences of a fuel

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

considering the degradation of the Boraflex in the spent fuel pool

racks and the presence of soluble boron.

Fuel Misloading

There is no significant increase in the probability of the

accidental misloading of spent fuel assemblies into the spent fuel

pool racks when considering the degradation of the Boraflex in the

spent fuel pool racks and the presence of soluble boron in the pool

water for criticality control. Fuel assembly placement and storage

will continue to be controlled pursuant to approved fuel handling

procedures to ensure compliance with the Technical Specification

requirements. These procedures will be revised as needed to comply

with the revised requirements which would be imposed by the proposed

Technical Specification changes.

There is no increase in the consequences of the accidental

misloading of spent fuel assemblies into the spent fuel pool racks

because criticality analyses demonstrate that the pool will remain

subcritical following an accidental misloading if the pool contains

an adequate boron concentration. Current Technical Specification

3.7.14 will ensure that an adequate spent fuel pool boron

concentration is maintained in the McGuire spent fuel storage pools.

A McGuire Station UFSAR change will revise Chapter 16, ``Selected

Licensee Commitments'', to provide for adequate monitoring of the

remaining Boraflex in the spent fuel pool racks. If that monitoring

identifies further reductions in the Boraflex panels which would not

support the conclusions of the McGuire Criticality Analysis, then

the McGuire TS's and design bases would be revised as needed to

ensure that acceptable subcriticality are maintained in the McGuire

spent fuel storage pools.

Significant Change in Spent Fuel Pool Temperature

There is no significant increase in the probability of either

the loss of normal cooling to the spent fuel pool water or a

decrease in pool water temperature from a large emergency makeup

when considering the degradation of the Boraflex in the spent fuel

pool racks and the presence of soluble boron in the pool water for

subcriticality control since a high concentration of soluble boron

has always been maintained in the spent fuel pool water. Current

Technical Specification 3.7.14 will ensure that an adequate spent

fuel pool boron concentration is maintained in the McGuire spent

fuel storage pools.

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

increase in the temperature of the water passing through the stored

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

result in a decrease in reactivity when Boraflex neutron absorber

panels are present in the racks. However, since a reduction in the

amount of Boraflex present in the racks is considered, and the spent

fuel pool water has a high concentration of boron, a density

decrease causes a positive reactivity addition. However, the

additional negative reactivity provided by the current boron

concentration limit, above that provided by the concentration

required to maintain keff less than or equal to 0.95

(1170 ppm), will compensate for the increased reactivity which could

result from a loss of spent fuel pool cooling event. Because

adequate soluble boron will be maintained in the spent fuel pool

water, the consequences of a loss of normal cooling to the spent

fuel pool will not be increased. Current Technical Specification

3.7.14 will ensure that an adequate spent fuel pool boron

concentration is maintained in the McGuire spent fuel storage pools.

A decrease in pool water temperature from a large emergency

makeup causes an increase in water density that would result in an

increase in reactivity when Boraflex neutron absorber panels are

present in the racks. However, the additional negative reactivity

provided by the current boron concentration limit, above that

provided by the concentration required to maintain keff

less than or equal to 0.95 (1170 ppm), will compensate for the

increased reactivity which could result from a decrease in spent

fuel pool water temperature. Because adequate soluble boron will be

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

decrease in pool water temperature will not be increased. Current

Technical Specification 3.7.14 will ensure that an adequate spent

fuel pool boron concentration is maintained in the McGuire spent

fuel storage pools.

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

kind of accident from any previously evaluated?

No. Criticality accidents in the spent fuel pool are not new or

different types of accidents. They have been analyzed in Section

9.1.2.3 of the Updated Final Safety Analysis Report and in

Criticality Analysis reports associated with specific licensing

amendments for fuel enrichments up to 4.75 weight percent U-235.

Specific accidents considered and evaluated include fuel assembly

drop, accidental misloading of spent fuel assemblies into the spent

fuel pool racks, and significant changes in spent fuel pool water

temperature. The accident analysis in the Updated Final Safety

Analysis Report remains bounding.

The possibility for creating a new or different kind of accident

is not credible. The amendment proposes to take credit for the

soluble boron in the spent fuel pool water for reactivity control in

the spent fuel pool while maintaining the necessary margin of

safety. Because soluble boron has always been present in the spent

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

always been a possibility, however this accident was not considered

credible. For the proposed amendment, the spent fuel pool dilution

evaluation (Attachment 7) demonstrates that a dilution of the boron

concentration in the spent fuel pool water which could increase the

rack keff to greater than 0.95 (constituting a reduction

of the required margin to criticality) is not a credible event. The

requirement to maintain boron concentration in the spent fuel pool

water for reactivity control will have no effect on normal pool

operations and maintenance. There are no changes in equipment design

or in plant configuration. This new requirement will not result in

the installation of any new equipment or modification of any

existing equipment. Therefore, the proposed amendment will not

result in the possibility of a new or different kind of accident.

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

of safety?

No. The proposed Technical Specification changes and the

resulting spent fuel storage operating limits will provide adequate

safety margin to ensure that the stored fuel assembly array will

always remain subcritical. Those limits are based on a plant

[[Page 27319]]

specific criticality analysis (Attachment 6) based on the

``Westinghouse Spent Fuel Rack Criticality Analysis Methodology''

described in Reference 1. The Westinghouse methodology for taking

credit for soluble boron in the spent fuel pool has been reviewed

and approved by the NRC (Reference 6). This methodology takes

partial credit for soluble boron in the spent fuel pool and requires

conformance with the following NRC Acceptance criteria for

preventing criticality outside the reactor:

(1) keff shall be less than 1.0 if fully flooded with

unborated water which includes an allowance for uncertainties at a

95% probability, 95% confidence (95/95) level; and

(2) keff shall be less than or equal to 0.95 if fully

flooded with borated water, which includes an allowance for

uncertainties at a 95/95 level.

The criticality analysis utilized credit for soluble boron to

ensure keff will be less than or equal to 0.95 under

normal circumstances, and storage configurations have been defined

using a 95/95 keff calculation to ensure that the spent

fuel rack keff will be less than 1.0 with no soluble

boron. Soluble boron credit is used to provide safety margin by

maintaining keff less than or equal to 0.95 including

uncertainties, tolerances and accident conditions in the presence of

spent fuel pool soluble boron. The loss of substantial amounts of

soluble boron from the spent fuel pool which could lead to exceeding

a keff of 0.95 has been evaluated (Attachment 7) and

shown to be not credible. Accordingly, the required margin to

criticality is not reduced.

The evaluations in Attachment 7, which show that the dilution of

the spent fuel pool boron concentration from the conservative

assumed initial boron concentration (2475 ppm) to the minimum boron

concentration required to maintain keff [less than or

equal to] 0.95 (440 ppm) is not credible, combined with the 95/95

calculation which shows that the spent fuel rack keff

will remain less than 1.0 when flooded with unborated water, provide

a level of safety comparable to the conservative criticality

analysis methodology required by References 2, 3 and 4.

Therefore the proposed changes in this license amendment will

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

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: J. Murray Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,

422 South Church Street, Charlotte, North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: April 6, 1999.

Description of amendment request: The proposed amendments would

expand the allowable values for Interlocks P-6 (Intermediate Range

Neutron Flux) and P-10 (Power Range Neutron Flux) in TS 3.3.1, Table

3.3.1-1, Function 16, Reactor Trip System Interlocks, as recommended by

Westinghouse.

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; 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.

Criterion 1--Would operation of the facility in accordance with

the requested amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

The reactor protection interlocks are provided to ensure reactor

trips are in the correct configuration for the current unit status.

They back up operator actions to ensure protection system functions

are not bypassed during unit conditions under which the safety

analysis assumes the functions are not bypassed. The proposed

changes involve changing the lower value of the P-10 permissive

(power range (PR) neutron flux) allowable values from [greater than

or equal to] 9% RTP to [greater than or equal to] 7% RTP, and

changing the P-6 permissive (intermediate range (IR) neutron flux)

allowable value from [greater than or equal to] 6E11 amp to [greater

than or equal to] 4E-11 amp. Changing the P-10 allowable value would

allow for tripping and resetting of the permissive at a lower

reactor power level. Changing the P-6 allowable value would allow

the source range (SR) channels to be blocked at a lower increasing

reactor power level and delay resetting of the permissive at a lower

decreasing reactor power level.

A review of the UFSAR Chapter 15 accident analyses determined

that no credit is taken for the SR reactor trip or the IR reactor

trip for any of the UFSAR accidents. Credit is taken for the PR low

setpoint trip for a feedwater system malfunction causing an increase

in feedwater flow accident (15.1.2), uncontrolled rod cluster

control assembly bank withdrawal from a subcritical or low power

startup condition accident (15.4.1), and spectrum of rod cluster

control assembly ejection accidents (15.4.8). All three of these

accident scenarios are bounded by cases at 0% RTP taking credit for

the PR low setpoint trip and cases at [greater than or equal to] 10%

RTP taking credit for the PR high setpoint trip. The uncontrolled

rod cluster control assembly bank withdrawal from power accident

(15.4.2) analyses are performed at initial power levels of 10%, 50%,

and 100% RTP to demonstrate that acceptable results are obtained for

a range of initial power levels. For this accident, the PR neutron

flux high setpoint trip, high pressurizer pressure trip, overpower

delta-T (OPDT) trip and overtemperature delta-T (OTDT) trip provide

core protection. With the P-10 reset function changed to as low as

7% RTP, the conclusions of Section 15.4.2 analysis would not change.

Since the uncontrolled bank withdrawal event is analyzed from both

zero power and 10% RTP, all low power initial conditions are

adequately bounded. Therefore, the proposed changes will not

increase the probability or consequences of an accident previously

evaluated.

Criterion 2--Would operation of the facility in accordance with

the requested amendment create the possibility of a new or different

kind of accident from any previously evaluated?

The proposed changes to the allowable values will provide

adequate deadbands between the trip and reset setpoints as well as

adequate margin for instrument drift. The reactor trip system

overpower trips continue to perform their safety function as assumed

in safety analyses. Only the permissives (P-6 and P-10) for blocking

and unblocking of overpower reactor trips are changed. The proposed

changes will not invalidate any of the UFSAR accident analyses. The

proposed changes will not introduce any new failure modes.

Therefore, the proposed changes will not create the possibility of a

new or different kind of accident from any previously evaluated.

Criterion 3--Would operation of the facility in accordance with

the requested amendment involve a significant reduction in a margin

of safety?

The proposed changes involve lowering the Technical

Specification allowable values associated with the P-10 and P-6

permissives for blocking and unblocking of reactor overpower trips.

The lowering of these allowable values is not considered a

significant reduction since it is just enough to accommodate a

deadband recommended by Westinghouse and a margin for instrument

drift. The proposed changes will not invalidate any UFSAR Chapter 15

accident analyses. Therefore, the proposed changes will not involve

a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,

[[Page 27320]]

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,

422 South Church Street, Charlotte, North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: April 26, 1999.

Description of amendment request: The proposed amendments would

revise the Technical Specifications to provide a method for obtaining a

Nuclear Regulatory Commission review of (a) the analytical details

regarding a revised methodology for determining steam generator tube

loads following a main steam line break, and (b) the crediting of the

main steam line break detection and feedwater isolation instrumentation

as a means for providing runout protection for the turbine-driven

emergency feedwater pump.

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?

No. The proposed changes involve: (a) revising the methodology

utilized to determine steam generator tube loads following a main

steam line break (MSLB); and (b) utilizing the MSLB detection and

feedwater isolation instrumentation as an additional means of

providing runout protection of the turbine-driven emergency

feedwater (EFW) pump.

The revised methodology utilized to determine steam generator

tube loads following a MSLB is consistent with the methodology

utilized in the MSLB containment response analysis which has

received Nuclear Regulatory Commission (NRC) approval. The revised

MSLB analysis reaches the same conclusion as the original analysis

(i.e., steam generator tube integrity is maintained). The new

analysis takes into consideration the operation of the MSLB

detection and feedwater isolation instrumentation to terminate main

feedwater (MFW) flow and inhibit the auto-start of or auto-stop the

turbine-driven EFW pump. This instrumentation is QA-1, whereas the

Integrated Control System (ICS) is non-safety. Furthermore, the

revised MSLB analysis results in a greater temperature difference

between the steam generator tube and shell, thus, more conservative

steam generator tube loads than those identified in the original

MSLB analysis.

Also, in the event that the MSLB detection and feedwater

isolation instrumentation does not function properly, the non-safety

ICS is still available to maintain steam generator water level at

the post-trip minimum level as assumed in the original analysis.

Currently, operator action is the only credited means to protect

the turbine-driven RFW pump from runout. The MSLB detection and

feedwater isolation instrumentation provides an additional method to

protect the turbine-driven EFW pump from runout. Crediting the MSLB

detection and feedwater isolation instrumentation simply adds

defense in depth.

There are no physical changes to the plant structures, systems,

or components (SSCs) or operating procedures, nor are there any

changes to safety limits or set points. Also, no new radiological

release pathways are created.

Thus, the proposed change does not significantly increase the

consequences of an accident previously evaluated.

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

from the accidents previously evaluated?

No. The reanalysis of the steam generator tube loads following a

MSLB accident is limited to an accident that is already evaluated in

the UFSAR. The methodology is similar to the current analysis for

the MSLB containment response. The effects of the MSLB on steam

generator tube integrity are the same as in the original analysis--

tube integrity is maintained.

The revised analysis takes into consideration the operation of

the MSLB detection and feedwater isolation instrumentation, which

terminates MFW flow and inhibits the auto-start of or auto-stops the

turbine-driven EFW pump following a MSLB. As assumed in the original

analysis, the non-safety ICS will remain available to control steam

generator water level at the post-trip minimum level should a

malfunction occur in the MSLB detection and mitigation circuit.

Should this malfunction occur, the resulting tube stresses would

decrease relative to the revised analysis.

Crediting the MSLB detection and feedwater isolation

instrumentation as a means to protect the turbine-driven EFW pump

from runout simply adds defense in depth.

There are no physical changes to the plant SSCs or operating

procedures. There are no new hazardous materials or potential

missiles. It does not introduce the possibility of any new or

different malfunctions. No safety limits or set points are changed.

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

new or different kind of accident.

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

No. The reanalysis of the steam generator tube loads following a

MSLB accident is similar to the current analysis for the previously

NRC approved MSLB containment response. The conclusion of the

revised MSLB steam generator tube load analysis is the same as the

conclusion in the original analysis--steam generator tube integrity

is maintained.

Crediting the MSLB detection and feedwater isolation

instrumentation as a means to protect the turbine-driven EFW pump

from runout simply adds defense in depth.

There are no safety limit, set point, design parameters, or

operating procedure changes required. The integrity of the fuel

cladding, reactor coolant system, and containment are preserved.

Thus, the proposed change does not involve a significant

reduction in a margin of safety.

Duke has concluded based on the above information that there are

no significant hazards involved in this LAR.

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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

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

No. 1, Pope County, Arkansas

Date of amendment request: April 9, 1999.

Description of amendment request: The proposed amendment would

revise the requirements affecting the surveillance methods for the

containment tendons, the conduct of containment visual inspections, and

the reporting methods employed in disseminating the results of these

inspections to the Nuclear Regulatory Commission.

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 proposed change to the ANO-1 [Arkansas Nuclear One, Unit 1]

TS [Technical Specifications] replaces previous requirements and

commitments to establish a containment inspection program based on

the guidance provided in Regulatory Guide 1.35, Revision 2 in favor

of regulations depicted in [Title] 10 [of the] CFR [Code of

[[Page 27321]]

Federal Regulations] 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix). ANO-1

is implementing a containment inspection program to comply with

these new regulatory requirements. The final rule specifies

requirements to assure that the critical areas of the containment

structure are routinely inspected to detect and take corrective

action for defects that could compromise structural integrity.

Maintaining reactor building structural integrity is independent

of the operation of the reactor coolant system (RCS), the reactor

protection system (RPS) and emergency core cooling system (ECCS).

The reactor building is not considered to be the initiator of any

accident previously evaluated. The physical location of inspection

details does not prevent or inhibit the reactor building from

functioning as designed to provide an acceptable barrier against

release of radioactive materials to the environment. Through

appropriate inspections and implementation of corrective actions for

any degradation discovered during the inspections that might lead to

containment structural failures, the probability or consequences of

accidents will not be increased.

Therefore, the removal of inspection details from the TS 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.

Maintaining containment structural integrity is independent of

the operation of the RCS, the RPS and ECCS. The proposed changes do

not change the design, configuration, or method of operation of the

plant. By implementing corrective actions for any degradation

discovered during the required inspections of the containment, the

possibility of a new or different kind of accident will not be

created. Implementation of the requirements of Subsection IWL of the

ASME [American Society of Mechanical Engineers] code and those of 10

CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) provide an equally

acceptable containment inspection program.

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 removal of the level of detail currently found in the ANO-1

TS regarding reactor building inspections and incorporating the

applicable requirements of Subsection IWL of the ASME code and of 10

CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) into the ANO-1

containment inspection program has no impact on any safety analysis

assumptions. Requirements associated with containment inspections

are controlled by safety related procedure 5220.011. Sufficient

controls exist under the procedure change process at ANO-1 to ensure

current and future regulations and commitments are properly

addressed when making revisions to the containment inspection

procedure. The addition of structural integrity requirements to ANO-

1 TS Specification 3.6.1 imposes consistent requirements with those

previously specified in the ANO-1 TSs. The containment inspection

program ensures that the containment will function as designed to

provide an acceptable barrier against release of radioactive

materials to the environment. Through the implementation of the

containment inspection program, the existing margin of safety is

preserved.

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, Arkansas 72801.

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

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

NRC Section Chief: Robert A. Gramm.

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

No. 1, Pope County, Arkansas

Date of amendment request: April 9, 1999.

Description of amendment request: The proposed amendment would

revise the requirements associated with the station batteries and the

direct current (dc) sources to the 125 volt dc switchyard distribution

system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The switchyard 125V DC control power source requirements do not

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

evaluated with respect to the selection criteria of [Title] 10 [of

the] CFR [Code of Federal Regulations] 50.36. These control power

sources are not assumed to mitigate accident or transient events.

The effects of a loss of these control power sources are enveloped

by the Loss of Offsite Power (LOOP) event and relocation is

considered to have a non-significant impact on the probability or

severity of a LOOP event. These requirements will be relocated from

the TSs to an appropriate administratively controlled document and

maintained pursuant to 10 CFR 50.59.

Proposed changes incorporating the requirements of TS 3.7.1.D,

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

subsystems, in the new TS 3.7.3 results in a more stringent

requirement for the ANO-1 [Arkansas Nuclear One, Unit 1] TSs in that

reductions to lower conditions of operation in shorter periods of

time are now required. These more stringent requirements are not

assumed to be initiators of any analyzed events and will not alter

assumptions relative to mitigation of accident or transient events.

The proposed addition of TS 3.7.4 allowing continued operation

for a limited period of time with battery cell parameters not within

limits under certain conditions clarifies an allowance that

currently exists in the ANO-1 TS due to the absence of acceptance

criteria for the battery cell parameter surveillances.

Proposed changes in Surveillance Requirements and Frequencies

reflect current industry guidance on maintenance and testing of the

station batteries. These requirements, in themselves, are not

considered to be initiators of any analyzed accident condition.

Although some frequencies have been extended, continued performance

of maintenance activities in accordance with IEEE-450 [Institute of

Electrical and Electronic Engineers, ``Recommended Practice for

Maintenance Testing and Replacement of Vented Lead-Acid Batteries

for Stationary Applications], in addition to the required

Surveillance Requirements, ensures that corrective maintenance can

be performed prior to a condition challenging an operability limit.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The proposed changes revise the surveillance requirements, and

required actions associated with the 125VDC distribution system and

the battery cell parameters. The requirements associated with the

ANO-1 switchyard DC sources have been relocated to licensee control.

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

method of operation of the plant.

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

or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the

Margin of Safety.

Relocation of the switchyard 125V DC control power source

requirements has no impact on any safety analysis assumptions. In

addition, the requirements associated with these control power

sources are relocated to an owner controlled document for which

future changes will be evaluated pursuant to the requirements of 10

CFR 50.59.

Proposed changes incorporating the requirements of TS 3.7.1.D,

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

subsystems, in the new TS 3.7.3 impose more stringent requirements

than previously specified for ANO-1.

The proposed addition of TS 3.7.4 allowing continued operation

for a limited period of time with battery cell parameters not within

limits under certain conditions clarifies an allowance that

currently exists in the ANO-1 TS due to the absence of acceptance

criteria for the battery cell parameter surveillances.

[[Page 27322]]

Proposed changes in Surveillance Requirements and Frequencies

reflect current industry guidance on maintenance and testing of the

station batteries. Although some frequencies have been extended,

continued performance of maintenance activities in accordance with

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

ensures that corrective maintenance can be performed prior to a

condition challenging an operability limit.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

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

NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: March 17, 1999.

Description of amendment request: The proposed amendment changes

the Perry Nuclear Power Plant as described in the Updated Safety

Analysis Report. The change incorporates a leak-off line in the

residual heat removal system. The leak-off line is designed to

eliminate an operator work around, which will significantly reduce the

collective dose to plant operations personnel.

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 modification has been described, and will be

procured and installed in accordance with the original design codes

and standards. The safety functions of the RHR [residual heat

removal] system have not been impacted by the change. Systems

supporting the operation of the RHR system have not been affected by

this modification. Though the modification affects the Containment

System, the containment remains capable of performing its associated

safety functions to the same level as the original design.

The accidents of concern are the Loss-Of-Coolant (LOCA) and the

Loss of Shutdown Cooling. The proposed change has been designed in

accordance with the original codes and standards. The proposed

change will not alter the operation of any plant equipment assumed

to function in response to the aforementioned analyzed events or

otherwise increase their failure probability. Therefore, the

probability of occurrence or the consequences of an accident

previously evaluated remains unchanged.

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

or different kind of accident from any previously evaluated.

The proposed modification has been designed, and will be

procured and installed in accordance with the original RHR system

design codes and standards. RHR system functions have not been

impacted by the change. Systems supporting the operation of the RHR

system have not been affected. Failure of the modification to

perform its design function due to leak-off line failure or blockage

would be identical to the current RHR system performance. Improper

operation of the valves associated with the modification have been

evaluated and will not prevent or otherwise inhibit the RHR or

Containment systems from performing their applicable safety

functions.

Missile generation is not a concern since no mechanisms

conducive to missile generation have been introduced. Electrical

analyses have shown there is no adverse effect upon the diesel

generator loadings. A single failure of the new configuration will

not result in more than the loss of a single RHR loop which is

already analyzed. Therefore, the possibility of a new or different

kind of accident from any previously evaluated has not been created.

3. The proposed change will not involve a significant reduction

in the margin of safety.

The proposed modification has been designed, and will be

procured and installed in accordance with the original RHR system

design codes and standards. The RHR and Containment systems remain

capable of performing their safety functions. Systems supporting the

operation of the RHR system have not been affected. Hence, the RHR

system margin of safety with respect to safety classification,

protection, redundancy, and seismic classification remains

unaffected.

The margins of safety contained in the Technical Specifications

and the associated Bases also remain unaffected by this

modification. Specifically, Technical Specifications 3.4.6,

``Reactor Coolant System Pressure Isolation Valve Leakage'; 3.4.9,

``RHR Shutdown Cooling System--Hot Shutdown'; 3.4.10, ``RHR Shutdown

Cooling System--Cold Shutdown'; 3.6.2.1, ``Suppression Pool Average

Temperature'; and 3.6.2.2, ``Suppression Pool Water Level'; and the

associated Bases remain unchanged and fully applicable. Hence, the

margins of safety defined in the Technical Specifications remains

unaffected.

Therefore, the proposed modification 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: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

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

NRC Section Chief: Anthony J. Mendiola.

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

Unit No. 1, Washington County, Nebraska

Date of amendment request: March 31, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to (1) increase the minimum reactor

coolant system (RCS) flow rate limit, (2) delete the reactor coolant

flow rate footnote, and (3) change the minimum frequency surveillance

for RCS flow rate.

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.

Combustion Engineering (ABB/CE) in Thermal-Hydraulic Report CR-

94-19-CSE95-1131, Revision 0 performed a comprehensive evaluation of

the effects the removal of the orifice plates would have on steam

generator tube degradation. It was concluded that the removal of the

orifice plates would increase the primary flow rate by approximately

5%.

The removal of the orifice plates was estimated to increase the

probability of tubes requiring repair over the lifetime of the

plant. However, the presence of the orifice plates had prevented

inspection of approximately 22% of the steam generator tubes for

circumferential cracks on the hot-leg side. Therefore, it was

concluded that the removal of the orifice plates did not increase

the probability of steam generator tube failure, given that the

tubes previously covered by the plates are now inspected each outage

in accordance with the Electrical Power Research Institute

Pressurized Water Reactor (EPRI PWR) steam generator examination

guidelines. Fort Calhoun Station is using the eddy current

inspection technology to ensure that tubes showing evidence of a

crack exceeding the present plugging criteria will be repaired or

removed from service. Industry experience has shown that even in

cases of severely degraded tubes, the

[[Page 27323]]

resulting primary to secondary leak rates are insignificant compared

to those analyzed in the design basis steam generator tube rupture

event.

Calculation of the Reactor Coolant Flow Rate using the heat

balance methodology once every refueling outage is consistent with

requirements contained in the NUREG 1432, Improved Technical

Specifications for Combustion Engineering Plants' surveillance

requirement 3.4.1.4.

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

or different kind of accident from any accident previously

evaluated.

The original orifice plates were installed on each steam

generator hot leg tube sheet in the primary inlet plenum as a field

modification prior to the initial fuel load in the year 1973. The

orifice plates were designed to increase the hydraulic resistance of

the primary coolant flow rate in the associated tubes, thereby

reducing the primary coolant temperature inside the tubes. Reduction

of the primary coolant temperature and flow rate would decrease the

heat flux, thus improving the steam quality and reducing the

potential for dry-out and surface deposits on the outer surface of

the tubes. However, due to inaccessibility, these originally

installed orifice plates had prevented tube inspection in the hot

leg tube sheet area, even with the latest state-of-the-art eddy

current probe technology. The orifice plates also prevented normal

repair techniques such as steam generator tube plugging and

sleeving.

The original orifice plates were removed during the 1996

refueling outage. However, there were concerns related to

Westinghouse fuel failures as a result of flow-induced vibration. To

address those concerns, new ``removable'' orifice plates were

installed to maintain the RCS flow rate at the previous level. Since

then, the remaining batches of the Westinghouse fuel considered most

susceptible to flow-induced vibration were replaced during the 1998

refueling outage, thus minimizing the concerns and allowing the

permanent removal of the ``removable'' orifice plates.

The removal of the ``removable'' orifice plates returned the

steam generators to their original design configuration. RCS flow

rate has increased by virtue of decreased hydraulic resistance

through the steam generators. No other systems or components other

than the steam generators have been affected. The resulting change

in operational parameters (decreased reactor coolant Thot

temperature and increased flow rate) has been evaluated for the

Updated Safety Analysis Report Chapter 14. Potential adverse

consequences of the modifications were (1) increase in reactor

vessel component vibration, (2) increase in hydraulic loading, and

(3) increase in steam generator tube degradation for row 1-18 tubes.

The potential adverse consequences were evaluated and found to be

acceptable.

Calculation of the Reactor Coolant Flow Rate using the heat

balance methodology once every refueling outage is consistent with

requirements contained in the NUREG 1432, Improved Technical

Specifications for Combustion Engineering Plants' surveillance

requirement 3.4.1.4.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The removal of the orifice plates has resulted in approximately

a 5% increase in the reactor coolant flow rate. This has increased

the margin for minimum reactor coolant system flow rate specified in

Technical Specifications Section 2.10.4, Power Distribution Limits,

Item (5), DNBR Margin During Power Operation Above 15% of Rated

Power. Steam Generator tube inspections performed in accordance with

Technical Specifications Section 3.17, Steam Generator Tubes, have

not been adversely affected.

The increased flow rate has been analyzed for the thermal

hydraulic effects on the reactor core and was found acceptable.

Calculation of the Reactor Coolant Flow Rate using the heat

balance methodology once every refueling outage is consistent with

requirements contained in the NUREG 1432 [Improved Technical

Specifications for Combustion Engineering Plants] surveillance

requirement 3.4.1.4.

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

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

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

amendment request involves no significant hazards consideration.

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

South 15th Street, Omaha, Nebraska 68102.

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

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

NRC Project Director: Stuart A. Richards.

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

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

Date of amendment request: January 28, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

remove two lists of Containment Isolation Valves (CIVs) in Tables 3.6-1

and 4.4-1 and make related changes to TSs 1.10, 3.6.A.1, and 4.4 and

the associated bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response: No. Operation of Indian Point 3 in accordance with the

proposed license amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The removal of the two component listings (i.e., Tables

3.6-1 and 4.4-1) and the TS references to them from the TS requested

by this submittal is performed in accordance with the guidance

provided by the NRC in GL 91-08 [Generic Letter 91-08]. As

established by the NRC, in the aforementioned GL, such a change will

not alter existing TS requirements or those components to which they

apply. Required information contained in the two tables being

removed is duplicated in the FSAR [final safety analysis report] and

other appropriate plant procedures. Any subsequent changes regarding

the individual components (i.e., the containment isolation valves)

or their operation (e.g., valve positioning under administrative

controls) would be addressed in accordance with the requirements

specified in the Administrative Controls section of the TS regarding

changes to plant procedures and/or changes to the FSAR (i.e., 10 CFR

50.59). These changes will not alter any structure, system, or

component and, therefore, will not result in the possibility of an

increase in [the] probability or consequence of an accident

previously evaluated.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response: No. The proposed changes do not create the possibility

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

evaluated. The deletion of two component listings (i.e., Tables 3.6-

1 and 4.4-1) and the TS references to them from the Technical

Specifications and the removal of all references made in the TS

regarding these two listings will not alter how the individual

components (i.e.--the containment isolation valves) identified in

the tables are designed, operated, tested, or maintained. Testing of

CIVs will be performed as required by 10 CFR part 50, Appendix J and

IP3 TS 6.14.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: No. The proposed license amendment does not involve a

significant reduction in a margin of safety. The proposed changes

are in accordance with recommendations provided by NRC in Generic

Letter 91-08 and the Standard Technical Specifications, NUREG 1431.

These changes will maintain current safety margins while reducing

the regulatory/administrative burdens to both the NRC and to the

Power Authority. As stated, the changes will not result in changes

to the design, operation, or maintenance of the ClVs, and the

testing of the CIVs will be in accordance with 10 CFR 50 Appendix J

and IP3 TS 6.14.

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

[[Page 27324]]

involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: April 12, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

remove the footnote restriction found on page 3.1-36 which states that

the departure from nucleate boiling (DNB) analysis contains adequate

margin for Cycle 10, but needs to be reviewed/approved prior to Cycle

11.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: The proposed change does not involve a significant

increase in the probability or consequences of an accident

previously analyzed. The removal of the footnote on TS page 3.1-36

is an administrative change in that it does not affect the DNB

limits of the current TS. The footnote was added to the TS as part

of Amendment 175, which permitted the use of V+ fuel at IP3. The

footnote required the Authority to demonstrate that sufficient DNB

margin existed for Cycle 11, prior to achieving criticality for that

cycle. The NRC requested this DNB limitation because the

applicability of the WRB-1 correlation to predict DNB performance

for the V+ fuel had not been adequately proven by fuel tests.

Westinghouse has completed fuel tests which verify that the use of

the WRB-1 correlation with the 15 x 15 V+ fuel is conservative.

Therefore, this DNB limitation is no longer applicable and the

footnote can be removed.

2. Does the proposed license amendment create the possibility of

a new or different kind of accident from any accident previously

evaluated?

Response: The proposed change does not create the possibility of

a new or different kind of accident, as the removal of the footnote

on TS page 3.1-36 does not affect the current TS DNB limits, plant

equipment, or the way the plant is operated. This footnote was

inserted into the TS as part of Amendment 175, which permitted the

use of 15 x 15 V+ fuel at IP3. Westinghouse had used scaling

techniques to demonstrate that the WRB-1 correlation correctly

predicted the critical heat flux performance of the 15 x 15 V+

fuel. Since no fuel tests had been performed on this fuel design,

the NRC was concerned that the use of this correlation may be

unconservative. Therefore, approval to use the V+ fuel at IP3 was

granted based upon the DNB margin available during Cycle 10. This

limitation was contained in the footnote on TS page 3.1-36.

Westinghouse has recently completed fuel tests on 15 x 15 V+ fuel

which verify that the use of the WRB-1 correlation is conservative.

Therefore, the use of V+ fuel at IP3 is no longer dependent on the

amount of DNB margin available and the footnote can be removed.

3. Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: The proposed deletion of the footnote on TS page 3.1-

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

The footnote was introduced as part of Amendment 175, which

permitted the use of V+ fuel at IP3. The footnote required the

Authority to demonstrate that sufficient DNB margin existed for

Cycle 11, prior to achieving criticality for that cycle. The NRC

requested this DNB limitation because the applicability of the WRB-1

correlation to predict DNB performance for the V+ fuel had not been

adequately proven by fuel tests. Westinghouse has completed fuel

tests which verify that the use of the WRB-1 correlation with the 15

x 15 V+ fuel is conservative. Therefore, this DNB limitation is no

longer applicable and the footnote can be removed. The removal of

the footnote is an administrative change as deleting it does not

alter the current DNB margin or future DNB margins.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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

Generating Station, Salem County, New Jersey

Date of amendment request: March 29, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) by relocating the procedural

details of the Radiological Effluent Technical Specifications (RETS) to

the Offsite Dose Calculation Manual (ODCM). The TSs would also be

revised to relocate procedural details associated with solid

radioactive wastes to the Process Control Program (PCP). In addition,

the Administrative Controls section of the TSs would be revised to

incorporate programmatic controls for radioactive effluents and

environmental monitoring. The proposed changes are consistent with the

guidance provided in Generic Letter 89-01, ``Implementation of

Programmatic Controls for Radiological Effluent Technical

Specifications in the Administrative Controls Section of the Technical

Specifications and the Relocation of Procedural Details of RETS to the

Offsite Dose Calculation Manual or to the Process Control Program.''

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 do not affect accident initiators or

precursors and do not alter the design assumptions, conditions,

configuration of the facility or the manner in which the plant is

operated. The proposed changes do not alter or prevent the ability

of structures, systems, or components to perform their intended

function to mitigate the consequences of an initiating event within

the acceptance limits assumed in the Updated Final Safety Analysis

Report (UFSAR). The proposed changes are administrative in nature

and do not change the level of programmatic controls and procedural

details relative to radiological effluents.

Implementation of programmatic controls for RETS in TS will

assure that the applicable regulatory requirements pertaining to the

control of radioactive effluents will continue to be maintained.

Since there are no changes to previous accident analysis, the

radiological consequences associated with these analyses remain

unchanged, therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. The proposed changes have no impact on component

or system interactions. The proposed changes are administrative in

nature and do not change the level of programmatic controls and

procedural details relative to radiological

[[Page 27325]]

effluents. Therefore, these changes will not create the possibility

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

evaluated.

(3) The proposed change does not involve a significant reduction

in a margin of safety.

There is no impact on equipment design or operation and there

are no changes being made to the TS required safety limits or safety

system settings that would adversely affect plant safety as a result

of the proposed changes. The proposed changes are administrative in

nature and do not change the level of programmatic controls and

procedural details relative to radiological effluents. A comparable

level of administrative control will continue to be applied to those

design conditions and associated surveillances being relocated to

the ODCM or PCP. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 29, 1999 (TS 99-04).

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) for Sequoyah (SQN) Units 1 and

2 by deleting the Auxiliary Feedwater (AFW) suction pressure low

channel functional surveillance test. The licensee's analysis of the

performance history revealed that the monthly functional test of this

instrument channel does not provide an increased assurance of

operability that justifies the monthly 7 hours per unit system

unavailability that it creates.

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:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The probability of occurrence or the consequences for an

accident is not increased by this request. The proposal to delete

the monthly channel functional test for the auxiliary feedwater

(AFW) suction pressure low functions does not alter the way any

structure, system or component functions, does not modify the manner

in which the plant is operated, and reduces equipment out-of-service

time. This request does not degrade the ability of AFW to perform

its intended function. Therefore, the pressure switches will be

available to perform their intended function.

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

new or different kind of accident from any accident previously

evaluated.

A possibility for an accident or malfunction of a different type

than any evaluated previously in SQN's FSAR [Final Safety Analysis

Report] is not created. The proposal does not alter the way any

structure, system or component functions and does not modify the

manner in which the plant is operated. Therefore, the possibility of

a new or different kind of accident previously evaluated is not

created by the proposed change to delete the monthly functional test

of the AFW pressure switches.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The margin of safety has not been reduced since the test

methodologies are not being changed. Increasing the surveillance

interval does not change the results of accident analysis by this

request. The proposed change to delete the AFW system pressure low

channel functional test does not involve a significant reduction in

the margin of safety. The new frequency will not reduce the

reliability of the system and increases overall system availability.

Therefore, changing the frequency of the surveillance 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 3740.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902.

NRC Section Chief: Sheri Peterson.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 29, 1999 (TS 99-03).

Description of amendment request: The proposed amendment would add

new actions to Technical Specification (TS) Limiting Condition for

Operations (LCOs) 3.3.3.1 and 3.7.7 to address the situation when one

channel of radiation monitoring control room emergency ventilation

system actuation equipment is inoperable and would expand the mode of

applicability for LCOs 3.3.3.1 and 3.7.7 to include periods when

movement of irradiated fuel assemblies are involved and defines actions

to take in these instances.

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:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed revision does not change any plant functions or

equipment operating practices for the radiation monitoring system

and control room emergency ventilation system (CREVS). The radiation

monitoring instruments and the CREVS are not considered to be the

source of any accident evaluated in the Final Safety Analysis

Report. These features provide accident mitigation functions that

will be utilized in response to postulated accident conditions. The

activities and failures that could contribute to the initiation of

an accident are not affected by the implementation of this revision.

This revision provides for more stringent requirements for operation

of the facility (additional limiting condition for operation [LCO]

actions and applicability requirements). Therefore the proposed

activity will not increase the probability of an accident.

The proposed activity does not affect accident mitigation

capabilities or the radiation release amounts for postulated

accidents. This TS change will not affect requirements that the

radiation monitoring system and CREVS be maintained to support

accident mitigation. The functions and testing will remain the same

while operability requirements will become more stringent. This TS

change enhances the requirements associated with CREVS and the

initiation of this system such that inoperabilities are

appropriately handled to reduce the safety impact of component

inoperabilities. Therefore, the proposed change will not increase

the consequences of an accident and could reduce the consequences by

limiting operation with inoperable components and requiring the

application of appropriate actions for all conditions that could

result in a postulated accident that CREVS was designed to mitigate.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change provides more stringent operating

requirements for operation of the facility. The proposed

[[Page 27326]]

activity will not change any plant function or operating practice

that could impact accident initiators. Therefore, these more

stringent requirements do not result in operation that will increase

the probability of any postulated accidents. In addition, CREVS and

the associated actuation features are not considered to be the

source of an accident. Therefore, the proposed activity will not

create the possibility of an accident of a different kind.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed activity does not impact plant setpoints designed

to maintain the assumptions in the safety analysis or limits for the

actuation of systems to mitigate accidents. Plant functions and

operating practices will not be altered by the implementation of

more stringent requirements for operation of the facility. These

requirements, by definition, provide additional restrictions to

enhance plant safety. Therefore, the proposed activity will not

reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902.

NRC Section Chief: Sheri Peterson.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: February 1, 1999, as supplemented on

April 19 and April 23, 1999.

Description of amendment request: The amendment request proposes a

total replacement of current Technical Specifications Section 6,

``Administrative Controls.'' Administrative changes to certain other

sections of Technical Specifications are also being made to conform to

the changes resulting from the re-write of Section 6.

The proposed changes represent a comprehensive upgrade of Section 6

of the Vermont Yankee Technical Specifications, incorporating

improvements in content and format based on industry standards. In

accordance with industry practice some Technical Specifications

requirements are being relocated to the recently implemented Vermont

Yankee Technical Requirements Manual (TRM), Offsite Dose Calculation

Manual (ODCM), or Vermont Yankee Operational Quality Assurance Manual

(VOQAM) and will be eliminated from the Technical Specification upon

NRC approval.

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 proposed changes have no effect on plant hardware, plant

design, safety limit setting, or plant system operation and

therefore do not modify or add any initiating parameters that would

significantly increase the probability or consequences of an

accident previously evaluated.

No new modes of operation are introduced by the proposed changes

such that additional adverse consequences would result. Accordingly,

the consequences of previously analyzed accidents are not

deleteriously affected by this proposed license amendment.

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

from any accident previously evaluated, because:

The proposed changes do not involve any physical alteration of

the plant (no new or different type of equipment will be installed)

or any change in the methods governing normal plant operation. These

changes do not affect the operation of any systems or components,

nor do they involve any potential initiating events that would

create any new or different kind of accident. Therefore, the

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

kind of accident from any accident previously evaluated for VYNPS.

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

because:

The proposed changes have no impact on any safety analysis

assumptions. Consequently, no margin of safety as described in the

Final Safety Analysis Report and defined in the basis of any

Technical Specification is reduced as a result of these changes.

These proposed changes do not detrimentally affect the ability

of structures, systems and components important to safety to fulfill

their intended safety functions. Therefore, it is concluded that the

proposed changes do no[t] involve a significant reduction in a

margin of safety.

Additional Safety Considerations for Specific Changes Deemed to be

``Less Restrictive''

In accordance with the criteria set forth in 10 CFR 50.92, Vermont

Yankee has evaluated the proposed changes to the [Vermont Yankee

Nuclear Power Station] VYNPS Technical Specifications and determined

that they do not involve a significant hazards consideration. Those

changes which are deemed to be ``less restrictive'' have been subject

to the following additional consideration:

(a) Changes which are deemed to be ``less restrictive'' based

solely upon removal from the Technical Specifications and relocated

in VYNPC-controlled documents:

NRC's Technical Specifications Branch has conducted reviews of

the Administrative Controls section of Standard Technical

Specifications and concluded that certain provisions historically

contained in Technical Specifications can be relocated to other

licensee documents for which changes to those provisions are

adequately controlled by other regulatory requirements. In general,

Administrative Controls are those requirements not covered by other

Technical Specifications, but are considered necessary to assure

operation of the facility in a safe manner. Application of this

criterion can be based on two categories or requirements: (a)

requirements not covered by other regulatory requirements, but are

considered necessary to assure the safe operation of the facility or

(b) specific requirements that are broadly covered by regulations or

other regulatory controls, for which details need to be specified in

the Technical Specifications to ensure safe plant operation. In

general, however, Technical Specifications need not duplicate other

regulatory requirements.

As identified in Attachment A hereto, certain portions of the

current Technical Specifications are to be relocated to the

Technical Requirements Manual (TRM), Offsite Dose Calculation Manual

(ODCM), or the Vermont Yankee Operational Quality Assurance Manual

(VOQAM) and removed from the Technical Specifications. As an initial

step in this process, the subject requirements are being duplicated

in the TRM, ODCM, or VOQAM. Removal from the Technical

Specifications will occur upon NRC approval. The ability to relocate

these requirements is based on regulations and standards that

contain these provisions such that duplication in the Technical

Specifications is not necessary.

[1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:]

The TRM is a[n] FSAR level document and is incorporated by

reference into the FSAR. Changes to the TRM will be strictly

controlled by the 10 CFR 50.59 process to ensure that proper reviews

are conducted. The relocation of requirements to the VYNPC-

controlled TRM will not diminish the effectiveness of compliance

withthe relocated provisions. Since any changes to the TRM 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 analyzed will be allowed. Therefore, these

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Changes to the ODCM are controlled by current Technical

Specifications and require the reporting to the NRC of changes to

the

[[Page 27327]]

ODCM with sufficient information to support the changes together

with appropriate analyses or evaluations justifying the changes. The

relocation of these details to the ODCM is thus acceptable

considering the controls provided by existing regulations and the

controls remaining in Technical Specifications for ODCM changes.

Therefore, these changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Relocation of the Technical Specification Administrative

Controls related to quality assurance from the Technical

Specifications to the VOQAM is consistent with the guidance provided

by the NRC in Administrative Letter 95-06, ``Relocation of Technical

Specification Administrative Controls Related to Quality

Assurance.'' Changes to the VOQAM are subject to the change control

process in 10CFR50.54(a). These provisions are adequate to ensure

that quality assurance program commitments are not reduced without

prior NRC approval. Therefore, these changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

accident from any accident previously evaluated, because:]

The proposed changes do not involve any physical alteration of

the plant (no new or different type of equipment will be installed)

or a change in the methods governing normal plant operation. The

proposed change will not impose or eliminate any requirements, and

adequate control of the information will be maintained. Thus, this

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

accident from any accident previously evaluated.

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

because:]

The proposed changes will not reduce a margin of safety because

they have no impact on any safety analysis assumption. In addition,

the details to be transposed from the Technical Specifications to

the TRM, ODCM, and VOQAM are the same as the existing Technical

Specifications. Since any future changes to these provisions in the

TRM will be evaluated per the requirements of 10CFR50.59 and

Technical Specifications already requires supporting information be

submitted to the NRC for ODCM changes, no reduction (significant or

insignificant) in a margin of safety will be allowed. The provisions

of 10CFR50.54(a) are adequate to control changes to the VOQAM and

maintain current margins of safety.

Based on 10CFR50.92, the existing requirement for NRC review and

approval of revisions (to the Technical Specifications provisions

proposed for relocation) does not have a specific margin of safety

upon which to evaluate. However, since the proposed changes are

consistent with industry standards, approved by the NRC, revising

the Technical Specifications to relocate these provisions will not

diminish administrative controls necessary to assure the safe

operation of the facility.

(b) Change [9] identified in Attachments A and D [of the

February 1, 1999, submittal]:

This change proposes to relax the requirement to have an

individual qualified in radiation protection procedures onsite at

all times. The proposed change will allow the position to be vacant

for up to two hours in order to provide for unexpected absence.

[1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:]

The proposed change does not affect the probability of an

accident. The actions of an individual qualified in radiation

protection procedures are not assumed to be an initiator of an

accident. Also, the consequences of an accident are not affected by

the presence of an individual qualified in radiation protection

procedures. This proposed change does not impact the assumptions of

any design basis accident. This change will not alter assumptions

relative to the mitigation of an accident or transient event. This

change will not have any impact on the safe operation of the plant

because the presence of a person qualified in radiation protection

procedures is not required for the mitigation of any accident.

Therefore, this change will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

accident from any accident previously evaluated, because:]

This change will not physically alter the plant (no new or

different type of equipment will be installed). The changes in

methods governing normal plant operation are consistent with the

current safety analysis assumptions. Therefore, this change will not

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

any accident previously evaluated.

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

because:]

The margin of safety is not affected by the presence or absence

onsite of an individual qualified in radiation protection

procedures. This proposed change has no effect on the assumptions of

any design basis accident. This change has no impact on the safe

operation of the plant since the presence onsite of an individual

qualified in radiation protection procedures is not required for the

mitigation of an accident. This change does not affect any plant

equipment or requirements for maintaining plant equipment. The

safety analysis assumptions will still be maintained, thus no

question of safety exists. Therefore, this change does not involve a

significant reduction in a margin of safety.

(c) Change [10] identified in Attachments A and D [of the

February 1, 1999, submittal]:

This change proposes to incorporate the allowances of a

temporary deviation from the shift staffing levels of

10CFR50.54(m)(2)(i) for up to two hours. In addition, this change

proposes to apply these same allowances to the positions of Shift

Engineer and non-licensed operators.

[1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:]

The proposed change does not affect the probability of an

accident. The shift staffing level requirements are not assumed to

be an initiator or any analyzed event. Also, the consequences of an

accident are not affected by these temporary deviations to the shift

staffing levels. This proposed change does not impact the

assumptions of any design basis accident. This change will not alter

assumptions relative to the mitigation of an accident or transient

event, since 10CFR50.54(m) (ii) and (iii) still maintain the

requirements for the presence of licensed operators and senior

operators. This change has no impact on the safe operation of the

plant. The level of shift staffing will still be maintained as

required by 10CFR50.54(m) (ii) and (iii) and does not affect any

plant equipment or requirements for maintaining plant equipment. The

temporary deviations from the shift staffing level for up to two

hours to provide for unexpected absence, provided immediate action

is taken to fill the required position is acceptable in terms of

staffing requirements for the mitigation of an accident due to the

low probability of an accident occurring during these short-term,

infrequent deviations and the remaining licensed operators and

senior operators. Therefore, this change will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

accident from any accident previously evaluated, because:]

This change will not physically alter the plant (no new or

different type of equipment will be installed). The temporary

deviations from shift staffing levels are consistent with the

current safety analysis assumptions. Therefore, this change will not

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

any accident previously evaluated.

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

because:]

The margin of safety in not reduced by allowing these temporary

deviations from shift staffing levels due to unforeseen events. This

proposed change has no effect on the assumptions of any design basis

accident. This change has no impact on the safe operation of the

plant since 10CFR50.54(m) (ii) and (iii) still maintain the

requirements for the minimum number of licensed operators and senior

operators necessary to safely operate the plant. This change does

not affect any plant equipment or requirements for maintaining plant

equipment. The safety analysis assumptions will still be maintained,

thus no question of safety exists. Therefore, this change does not

involve a significant reduction in a margin of safety.

(d) Changes [38] and [39] identified in Attachments A and D [of

the February 1, 1999, submittal]:

In accordance with 10CFR20.1601 (c), these changes propose

alternative methods for controlling access to high radiation areas

consistent with the intent of 10CFR20.1601 (a) and (b).

[1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:]

The proposed changes do not affect the probability of an

accident. The controls used for access to high radiation areas are

not assumed in the initiation of any analyzed event. Also, the

consequences of an accident are not affected by these changes. These

changes are both consistent with good

[[Page 27328]]

radiological practices and will provide an adequate level of

radiation protection. These proposed changes do not impact the

assumptions of any design basis accident. These changes will not

alter assumptions relative to the mitigation of an accident or

transient event. These changes have no impact on safe operation of

the plant. Therefore, these changes will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

accident from any accident previously evaluated, because:]

The proposed changes will not create the possibility of an

accident. These changes will not physically alter the plant (no new

or different type of equipment or system will be installed). The

changes in methods governing normal plant operations are consistent

with the current safety analysis assumptions and deal only with

personnel exposure to radiation, not reactor safety. Therefore,

these changes will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

because:]

The margin of safety is not reduced due to these proposed

changes. These changes are both consistent with good radiological

safety practice and have been found to provide adequate levels of

radiation protection. In addition, these changes provide the benefit

of ensuring radiation dose to workers can be minimized by providing

the flexibility to select the best means of providing access control

to a high radiation area, given the plant area and radiological

conditions. These proposed changes have no impact on the safe

operation of the plant. No change in analytic limits or setpoints is

introduced by these changes. The safety analysis assumptions will

still be maintained, thus no question of nuclear safety exits.

Therefore, these changes do not involve a significant reduction in a

margin of safety.

(e) Change [49] identified in Attachments A and D [of the

February 1, 1999, submittal]:

This change proposes to relax the requirement for submitting the

(now-named) Occupational Radiation Exposure Report from the

currently required date of March 1 to April 30 of each year. April

30 is now the industry standard date for submittal of such reports.

[1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:]

The proposed change does not affect the probability of an

accident. The submittal date of the Occupational Radiation Exposure

Report is not assumed to be an initiator of any analyzed event.

Also, the consequences of an accident are not affected by the

submittal date of this report. This proposed change does not impact

the assumptions of any design basis accident. This change will not

alter assumptions relative to the mitigation of an accident or

transient event. This change has no impact on the safe operation of

the plant. The report will still be required to be submitted each

year and does not affect any plant equipment or requirements for

maintaining plant equipment. The submittal date of this report is

not required for the mitigation of any accident. Therefore, this

change will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

accident from any accident previously evaluated, because:]

The proposed change will not create the possibility of an

accident. This change will not physically alter the plant (no new or

different type of equipment will be installed). The change in method

governing submittal of this report does not affect current safety

analysis assumptions. Therefore, this change will not create the

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

previously evaluated.

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

because:]

The margin of safety i[s] not reduced by allowing the report to

be submitted 60 days later. This proposed change has no effect on

the assumptions of the design basis accident. This change has no

impact on the safe operation of the plant. The report will still be

required to be submitted each year and does not affect any plant

equipment or requirements for maintaining plant equipment. The

safety analysis assumptions will still be maintained, thus no

question of safety exists. Therefore, this change does not involve a

significant reduction in a margin of safety.

(f) [Change [64] identified in Attachments A and D [of the

February 1, 1999, submittal]:

This change proposes to relax the requirement for submitting the

(now-named) Annual Radiological Environmental Operating Report from

the currently required date of May 1 to May 15 of each year. May 15

is now the industry standard date for submittal of such reports.

[1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:]

The proposed change does not affect the probability of an

accident. The submittal date of this report is not assumed to be an

initiator of any analyzed event. Also, the consequences of an

accident are not affected by the submittal date of this report. This

proposed change does not impact the assumptions of any design basis

accident. This change will not alter assumptions relative to the

mitigation of an accident or transient event. This change has no

impact on the safe operation of the plant. The report will still be

required to be submitted each year and does not affect any plant

equipment or requirements for maintaining plant equipment. The

submittal date of this report is not required for the mitigation of

any accident. Therefore, this change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

accident from any accident previously evaluated, because:]

The proposed change will not create the possibility of an

accident. This change will not physically alter the plant (no new or

different type of equipment will be installed). The change in method

governing submittal of this report does not affect current safety

analysis assumptions. Therefore, this change will not create the

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

previously evaluated.

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

because:]

The margin of safety i[s] not reduced by allowing the report to

be submitted 14 days later. This proposed change has no effect on

the assumptions of the design basis accident. This change has no

impact on the safe operation of the plant. The report will still be

required to be submitted each year and does not affect any plant

equipment or requirements for maintaining plant equipment. The

safety analysis assumptions will still be maintained, thus no

question of safety exists. Therefore, this change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: April 20, 1999.

Description of amendment request: The amendment request proposes

changes to the existing requirements associated with the unloading and

loading of fuel in the reactor vessel.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

VY has determined that the proposed change to reload the reactor

core in a spiral pattern beginning around a Source Range Monitor

(SRM) does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The design basis

accident associated with refueling is the Refueling Accident; i.e.,

the accidental dropping of a fuel bundle onto the top of the core.

There is no assumption as to

[[Page 27329]]

the core loading pattern in the analysis of this accident. The

analyzed abnormal operational transients associated with refueling

are: (1) the Control Rod Removal Error During Refueling, and (2) the

Fuel Assembly Insertion Error During Refueling. There is no

assumption as to the core loading pattern in the analyses of these

transients. The Fuel Assembly Insertion Error During Refueling

transient involves mislocated and rotated fuel assembly loading

errors. However, a change in the approved core loading pattern has

no impact on the probability of mislocating or rotating a bundle

while following that pattern. Furthermore, the proposed change

implements a core loading pattern that provides improved flux

monitoring as compared to the pattern prescribed by the current

Technical Specifications. When loading the core in accordance with

the proposed change, the SRM indication will be indicative of the

true flux of the loaded fuel, as the creation of flux traps

(moderator filled cavities surrounded on all sides by fuel) is

precluded.

The Technical Specification Bases are under the purview of

10CFR50.59. As such, subsequent changes made via 10CFR50.59 to the

information relocated to the Bases are not allowed to increase the

probability or consequences of an accident previously evaluated.

Therefore, relocating the details of the core loading pattern to the

Bases does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The SRMs and the core loading pattern are not initiators of any

accident previously evaluated. As such, the subject changes cannot

affect the probability of an accident previously evaluated. The core

loading pattern is not assumed in the mitigation of any accident.

Since the proposed change provides improved flux monitoring by the

SRMs, operators will have more accurate indication and SRM automatic

trip functions will actuate more accurately. As such, any event

mitigation function provided by the SRMs is enhanced by this change.

Therefore, the associated changes do not involve a significant

increase in the consequences of an accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

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

previously evaluated.

VY has determined that the proposed change does not create the

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

previously evaluated. VY proposes to change the core reloading and

offloading patterns to start and stop, respectively, at an SRM

versus the geometric center of the core as prescribed by current

Technical Specifications. This ensures that flux monitoring

instrumentation is always OPERABLE in the fueled region of the

vessel. There is no separation of the monitoring device from the

fuel by cavities of water as is the case with the pattern prescribed

by the current Technical Specifications. As such, flux monitoring is

enhanced during core reloading and offloading. This change is

conservative relative to the current requirements. Therefore, no new

categories or types of accidents are created.

Additionally, the Technical Specification Bases are under the

purview of 10CFR50.59. As such, subsequent changes made via

10CFR50.59 to the information relocated to the Bases are not allowed

to create the possibility for an accident or malfunction of a

different type than any evaluated previously in the safety analysis

report. Therefore, relocating the details of the core loading

pattern to the Bases does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

VY has determined that the proposed change does not involve a

significant reduction in a margin of safety. Loading around the

geometric center of the core as prescribed by the current Technical

Specifications results in cells of moderator separating the fuel

from the instrumentation monitoring its flux. This change requires

the flux monitoring instrumentation to be in the fueled region, and,

in so doing, provides for more accurate monitoring of core flux

during core reloading and offloading. As such, the operators will

have more accurate indication and SRM automatic trip functions will

actuate when the actual flux reaches the trip setpoints. This

corrects non-conservatisms that result from cells of moderator

separating the fuel from the instrumentation. Therefore, this change

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

Additionally, the details of the loading pattern are relocated

from the Technical Specifications to the Bases. Since any future

changes to the Bases will be evaluated per the requirements of 10

CFR 50.59, no reduction in a margin of safety will be allowed.

Therefore, relocating the core loading pattern details to the Bases

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: April 7, 1999.

Description of amendment request: The proposed amendment would

revise the minimum critical power ratio (MCPR) limit in Technical

Specification (TS) 2.1.1.2, for the ATRIUM-9X and the SVEA-96 fuel for

one and two recirculation loop operation. The proposed amendment would

add a new reference in TS 5.6.5, ``Core Operating Limits Report.'' The

reference cites ANFB Critical Power Correlation Uncertainty for Limited

Data Sets, ANF1125(P)(A), Supplement 1, Appendix D, Siemens Power

Corporation-Nuclear Division, July 1998.

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 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 systems designed to mitigate those

consequences. Limits have been established consistent with NRC

approved methods to ensure that fuel performance during normal,

transient, and accident conditions is acceptable. The proposed

Technical Specifications amendment uses conservatively established

SLMCPR [safety limit minimum critical power ratio] values for WNP-2

such that the fuel is protected during normal operation as well as

during plant transients or anticipated operational occurrences.

The probability of an evaluated accident is not increased by the

use of the ATRIUM-9X MCPR safety limit of 1.10 (two loop operation)

or 1.11 (single loop operation). The ATRIUM-9X fuel was evaluated by

SPC (Reference 5) [Letter KVW:98:148 dated July 8, 1998, KV Walters,

(Siemens Power Corporation), to RA Vopalensky (Supply System),

``MCPR Safety Limit Reanalysis for WNP-2 Cycle 11''] using the

additive constant uncertainty for ATRIUM-9X fuel of 0.0201 which is

contained in the NRC safety evaluation approval of Reference 4 [ANFB

Critical Power Correlation Uncertainty for Limited Data Sets, ANF-

1125(P)(A), Supplement 1, Appendix D, Siemens Power Corporation--

Nuclear Division, July 1998]. Based upon the NRC approved additive

constant of uncertainty of 0.0201, as documented in Reference 5, at

least 99.9% of the SPC ATRIUM-9X fuel rods would be expected to

avoid boiling transition with a SLMCPR of 1.10 during two loop

operation and 1.11 during single loop operation.

The probability of an evaluated accident is not increased by the

use of the ABB SVEA-96 SLMCPRs of 1.10 (two loop operation) or 1.12

(single loop operation). NRC approved

[[Page 27330]]

methodology documented in CENPD-300-P-A, ``Reference Safety Report

for Boiling Water Reactor Reload Fuel'', July 1996 (Reference 3) was

used in deriving these ABB SVEA-96 SLMCPR values. The ABB evaluation

as a function of cycle exposure established that late in Cycle 15

conservative two loop and single loop SLMCPRs of 1.10 and 1.12,

respectively, can be used to represent the entire cycle.

The SLMCPR changes do not require any physical plant

modifications, physically affect any plant component, or entail

changes in plant operation. Therefore, no individual precursors of

an accident are affected.

Since the operability of plant systems designed to mitigate any

consequences of accidents have not changed, the consequences of an

accident previously evaluated are not expected to increase.

2. The proposed change does not 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. This Technical Specification submittal

does not involve any modifications of the plant configuration or

allowable modes of operation. This Technical Specification change

establishes SLMCPRs for SPC fuel based upon the NRC approved

additive constant of uncertainty of 0.0201, as documented in

Reference 5. At least 99.9% of the SPC ATRIUM-9X fuel rods would be

expected to avoid boiling transition with an SLMCPR of 1.10 during

two loop operation or 1.11 during single loop operation.

Additionally, the ABB SVEA-96 SLMCPRs of 1.10 (two loop operation)

or 1.12 (single loop operation) were derived using the NRC approved

methodology documented in CENPD-300-P-P, ``Reference Safety Report

for Boiling Water Reactor Reload Fuel'', July 1996 (Reference 3).

Therefore, no new precursors of an accident are created and no new

or different kinds of accidents are created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Implementation of SLMCPRs derived by proven analytical methods

provides a margin of safety by ensuring that less than 0.1% of the

rods are expected to be in boiling transition if the MCPR limit is

not violated. Because the fuel design safety criteria of more than

99.9% of the fuel rods avoiding transition boiling during normal

operation as well as anticipated operational occurrences is met,

there is not 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 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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

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

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

NRC Project Director: Stuart Richards.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: April 20, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.4.11, ``RCS Pressure and

Temperature Limits,'' to update the curves that set forth the pressure

temperature limit lines. The curves provide the pressure temperature

limits for the operation of the reactor coolant system for heatup and

cooldown during inservice leak and hydrostatic testing, non-nuclear

heating and cooldown, and nuclear heating and cooldown.

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 pressure temperature shift is well within the operating

margins of plant equipment. Using the new non-nuclear and nuclear

heating and cooldown curves, higher temperature values for

corresponding pressures at temperatures which are closest to RT

NDT, further reduce the potential for brittle fracture.

The proposed 32 EFPY [effective full power years] curves were

developed using methodology that is consistent with the guidance in

Regulatory Guide 1.99, Revision 2, Appendix G of the ASME Code and

Appendix G of 10 CFR part 50. This methodology is recognized by the

NRC and the industry as providing acceptable margin.

Therefore, operation of WNP-2 in accordance with the proposed

amendment will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change has no impact on the previously analyzed

accidents or transients. The proposed change does not introduce any

credible mechanisms for unacceptable radiation release nor does it

require physical modification to the plant. The 32 EFPY curves are

calculated using a published methodology that was discussed with the

NRC.

The proposed change is also within any upper bound limit. The

only impact on plant operation is that the plant will be operated

with new pressure temperature limits derived from the proposed

alternative calculational methodology in place of the previously

approved model based on actual plant data.

Therefore, the operation of WNP-2 in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The results of testing reflected 30 ft-lb shifts and changes in

uppershelf energy of the base plate and the weld material. However,

the results are well within the values predicted by Regulatory Guide

1.99, Revision 2. Furthermore, the adjusted reference temperature

values and the upper shelf energy of the reactor beltline materials

are expected to remain within the limits of 10 CFR part 50, Appendix

G, for at least 32 effective full power years of reactor operation.

For the non-nuclear and nuclear heating and cooldown curves

(with a calculated through wall T), lower temperatures

which are closest to RTNDT, have an increased margin of

safety due to the higher required temperature values for a given

pressure than is required by current curve calculation methodology.

Thus additional margin to brittle fracture is achieved for non-

nuclear and nuclear heating.

Therefore, operation of WNP-2 in accordance with the proposed

amendment will 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 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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

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

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

NRC Project Director: Stuart Richards.

Previously Published Notices of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait

[[Page 27331]]

for this biweekly notice or because the action involved exigent

circumstances. They are repeated here because the biweekly notice lists

all amendments issued or proposed to be issued involving no significant

hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: December 31, 1997, as supplemented May

15, September 15, November 25, 1998 and January 28, 1999.

Description of amendment request: Revise the St. Lucie, Unit 2,

Technical Specifications to increase the capacity of the spent fuel

storage pool, in part, by allowing a credit for a certain soluble boron

concentration in the spent fuel pool.

Date of publication of individual notice in the Federal Register:

April 5, 1999 (64 FR 16502).

Expiration date of individual notice: May 5, 1999.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: April 19, 1999.

Brief description of amendments: The amendments would revise

Technical Specification Section 3/4.8.1.2, ``Electrical Power Systems,

Shutdown,'' and its associated bases to provide a one-time extension of

the 18-month surveillance interval for specific surveillance

requirements for Units 1 and 2. This surveillance will be performed

prior to the first entry into Mode 4 subsequent to receipt of the

requested T/S amendment. In addition, for Unit 2 only, a minor

administrative change is included to delete a reference to T/S 4.0.8,

which is no longer applicable. For Unit 1 only, an editorial change is

made to add the word ``or'' to action statement 3.8.1.2.

Date of publication of individual notice in Federal Register: April

29, 1999 (64 FR 23129).

Expiration date of individual notice: June 1, 1999.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch

Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of amendment request: April 6, 1999.

Description of amendment request: The proposed amendments would

allow an increase of 168 fuel assemblies in the storage capacity of

Unit 1's Spent Fuel Pool and an increase of 88 fuel assemblies in the

storage capacity of Unit 2's Spent Fuel Pool.

Date of publication of individual notice in Federal Register: May

4, 1999 (64 FR 23877).

Expiration date of individual notice: June 3, 1999.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

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

County, Illinois

Date of application for amendments: December 29, 1998.

Brief description of amendments: The amendments change Technical

Specification Tables 3.3.1-1 and 3.3.2-1 to revise the Allowable Values

for 12 functions of the Reactor Trip System and Engineered Safety

Features Actuation System.

Date of issuance: April 23, 1999.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 107, 107, 100 and 100.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9186). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 23, 1999.

No significant hazards consideration comments received: No.

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

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

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

Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN50-455,

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

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

County, Illinois

Date of application for amendments: October 30, 1998.

Brief description of amendments: The amendments revised the

Technical Specification (TS) requirements for

[[Page 27332]]

spent fuel pool inadvertent draindown elevation.

Date of issuance: May 3, 1999.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 108, 108 101, and 101.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 16, 1998 (63

FR 69335). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 3, 1999.

No significant hazards consideration comments received: No.

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

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

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

Wilmington, Illinois 0481.

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

Plant, Middlesex County, Connecticut

Date of application of amendment: June 2, 1998, and as supplemented

by letters dated January 18 and March 9, 1999.

Brief description of amendment: The amendment relocates

requirements related to seismic monitoring instrumentation from the

Technical Specifications to the Technical Requirements Manual.

Date of issuance: April 28, 1999.

Effective date: Immediately; and shall be implemented within 60

days of issuance.

Amendment No.: 194.

Facility Operating License No. DPR-61: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50936). The January 18 and March 9, 1999, supplements contained

revised TS pages to account for TS changes issued by the NRC since the

original June 2, 1998, submittal, pages from the Updated Final Safety

Analysis Report and TRM, which were revised to support the June 2,

1998, request, and additional clarifications. The supplemental

information did not change the staff's initial proposed no significant

hazards consideration determination or expand the scope of the original

notice. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 28, 1999.

No significant hazards consideration received: No

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: August 21, 1996, as supplemented

May 2, 1997.

Brief description of amendment: The amendment revised Section 3.3.G

(Hydrogen Recombiner System and Post-Accident Containment Venting

System), the basis for Section 3.3.G, and Section 4.4, Table 4.4-1

(Containment Isolation Valves). This change permits removal of the

existing flame-type hydrogen recombiners, its supporting equipment, and

replacement with passive autocatalytic recombiners.

Date of issuance: April 27, 1999.

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

30 days.

Amendment No.: 200.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: Janaury 29, 1997 (62 FR

4345). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 27, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment: September 3, 1997.

Brief description of amendment: The amendment revises TS 3.14,

Control Room Ventilation, to be consistent with NUREG-1432, Standard

Technical Specifications, Combustion Engineering Plants.

Date of issuance: May 6, 1999.

Effective date: May 6, 1999.

Amendment No.: 186.

Facility Operating License No. DPR-20: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 24, 1999 (64 FR

14281). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 6, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423-3698.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina

Date of application of amendments: March 1, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications by adding a Note to Improved Technical

Specification (ITS) 3.9, ``Refueling Operations,'' Subsection 3.9.3,

``Containment Penetrations,'' Limiting Condition for Operation 3.9.3.b,

to state that the emergency air lock door is not required to be closed

when it is sealed with the temporary cover plate.

Date of Issuance: April 28, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days of issuance.

Amendment Nos.: Unit 1-303; Unit 2-303; Unit 3-303.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: 64 FR 14282 (March 24,

1999). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 28, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: March 1, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications by changing the number of required channels

shown in TS Table 3.3.8-1, ``Post Accident Monitoring Instrumentation''

for the Reactor Coolant System Hot Leg Temperature function from ``2

per loop'' to ``2.''

Date of Issuance: April 28, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days of issuance.

Amendment Nos.: Unit 1-304; Unit 2-304; Unit 3-304

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 24, 1999 (64 FR

14281). The Commission's related evaluation of

[[Page 27333]]

the amendments is contained in a Safety Evaluation dated April 28,

1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: April 30, 1998.

Brief description of amendment: The amendment revises the single

largest post-accident load capable of being supplied by the diesel

generators and relocates this value to the Bases for Technical

Specification (TS) Surveillance 4.8.1.1.2.c.3. TS Surveillance

4.8.1.1.2.c.3 has been revised to refer to ``the single largest post-

accident load'' rather than a specific numerical value for diesel

generator load reject testing. This change is consistent with the

guidance provided in NUREG-1432 , ``Improved Standard Technical

Specifications for Combustion Engineering Plants.''

Date of issuance: April 21, 1999.

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

30 days from the date of issuance.

Amendment No.: 204.

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56241). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: July 21, 1995.

Brief description of amendment: The amendment extends the

expiration date of Operating License NPF-29 for Grand Gulf Nuclear

Station, Unit 1, from June 16, 2022, to November 1, 2024. The extended

date is 40 years from the date the full-power license was issued for

the plant on November 1, 1984.

Date of issuance: April 26, 1999.

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

30 days of issuance.

Amendment No: 137.

Facility Operating License No. NPF-29: Amendment revises Operating

License No. NPF-29.

Date of initial notice in Federal Register: August 16, 1995 (60 FR

42605). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

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

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: December 16, 1998.

Brief description of amendment: The amendment changes Technical

Specification (TS) Section 2.1.1.2, ``Reactor Core [Safety Limits],''

by revising the two recirculation loop Minimum Critical Power Ratio

(MCPR) limit from 1.13 to 1.12 and the single recirculation loop MCPR

limit from 1.14 to 1.13. The revised limits are required to address the

River Bend Cycle 9 core design and operation. The proposed TS changes

are scheduled to be implemented following refueling outage 8, currently

scheduled to begin in April 1999.

Date of issuance: April 27, 1999.

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

to the startup following refueling outage 8.

Amendment No.: 105.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9190). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 27, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 8, 1998, as supplemented April

15, 1999.

Brief description of amendment: The amendment implements the

Boiling Water Reactor Owners Group Enhanced Option I-A for the reactor

stability long-term solution to the neutronic and thermal hydraulic

instability that is documented in NEDO-32339, Revision 1, ``Reactor

Stability Long-Term Solution, Enhanced Option I-A.''

Date of issuance: May 5, 1999.

Effective date: As of the date of issuance and shall be implemented

during refueling outage 8.

Amendment No.: 106.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64112). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 5, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: October 30, 1998, as

supplemented March 31, 1999.

Brief description of amendment: The amendment proposed to revise

the Final Safety Analysis Report (FSAR) and associated Improved

Technical Specification (ITS) Bases to reflect changes in the

methodology for the B spent fuel pool criticality analysis. The

proposed change is necessary due to Boraflex degradation in the B spent

fuel pool storage racks.

Date of issuance: April 27, 1999.

Effective date: April 27, 1999.

Amendment No.: 175.

Facility Operating License No. DPR-72: Amendment approves changes

to the FSAR and ITS Bases.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71966). The supplemental letter dated March 31, 1999, did not change

the original no significant hazards consideration determination.

[[Page 27334]]

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 27, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: January 27, 1999.

Brief description of amendment: The change would allow a one-time

extension of approximately 2 months of the steam generator tube

inspection interval in order for the inspection to coincide with the

next planned refueling outage.

Date of issuance: May 5, 1999.

Effective date: May 5, 1999.

Amendment No.: 176.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 10, 1999 (64 FR

11962). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 5, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: December 31, 1997, as

supplemented May 15, 1998, September 15, 1998, November 25, 1998, and

January 25, 1998.

Brief description of amendment: This change modified the St. Lucie

Unit 2 Technical Specifications to increase the capacity of the spent

fuel storage pool, in part, by allowing a credit for a certain soluble

boron concentration in the spent fuel pool.

Date of Issuance: May 6, 1999.

Effective Date: Upon issuance of license amendment package with

implementation by the end of the next scheduled refueling outage,

currently scheduled for April of 2000.

Amendment No.: 101.

Facility Operating License No. NPF-16: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 11, 1998 (63

FR 6985) and December 16, 1998 (63 FR 69340). Following the receipt of

the supplement dated November 25, 1998, and the staff's subsequent no

significant hazards consideration determination (63 FR 69340), the

supplement dated January 28, 1999, contained clarifying information

that did not change the no significant hazards consideration

determination. An additional notice was required, in accordance with 10

CFR 2.1107, due to an oversight (64 FR 16502, April 5, 1999). An

environmental assessment has been published in the Federal Register (64

FR 23133, April 29, 1999). In that assessment, the Commission

determined that the issuance of this amendment will not result in any

environmental impacts other than those evaluated in the Final

Environmental Statement.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 6, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: February 24, 1999.

Brief description of amendments: The amendments changed Technical

Specification (TS) 3/4.7.4 to permit the option of monitoring the

ultimate heat sink temperature afer the intake cooling water (ICW)

pumps but before the component cooling water heat exchangers which is

considered to be equivalent to temperature monitoring before the ICW

pumps.

Date of issuance: May 5, 1999.

Effective date: May 5, 1999.

Amendment Nos.: 200 and 194.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the TS.

Date of initial notice in Federal Register: March 24, 1999 (64 FR

14282). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 5, 1999.

No significant hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: July 14, 1998

Brief description of amendment: The proposed amendment changed the

Technical Specifications to revise the liquid and gaseous release rate

limits to reflect revisions to 10 CFR Part 20, ``Standards for

Protection Against Radiation.''

Date of issuance: May 3, 1999.

Effective date: May 3, 1999, to be implemented within 30 days from

the date of issuance.

Amendment No.: 163.

Facility Operating License No. DPR-36: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 13, 1999 (64 FR

2249). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: September 30, 1997.

Brief description of amendment: The proposed amendment revises

portions of Facility Operating License No. DPR-36 to delete License

Conditions 2.B.6.c, 2.B.6.e, 2.B.6.f, 2.b.6.g, 2.b.7(a), and 2.B.7(b)

which are no longer applicable due to the permanently shutdown and

defueled condition of the Maine Yankee Atomic Power Station. Orders

dated May 23, 1980, August 29, 1980, and September 19, 1980, are

rescinded due to their being superseded by the equipment qualification

rule (10 CFR 50.49).

Date of issuance: May 5, 1999.

Effective date: May 5, 1999, and shall be implemented within 30

days from the date of issuance.

Amendment No.: 164.

Facility Operating License No. DPR-36: The amendment revised the

Operating License.

Date of initial notice in Federal Register: December 3, 1997 (62 FR

63978). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 5, 1999.

[[Page 27335]]

No significant hazards consideration comments received: No.

Local Public Document Room location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578.

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

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

California

Date of application for amendments: January 14, 1998, as

supplemented by letters dated May 19, 1998, September 28, 1998, and

three letters dated February 5, 1999.

Brief description of amendments: The amendments authorize revisions

to the licensing basis as described in the Final Safety Analysis Report

(FSAR) Update to incorporate the modification to the 230 kV offsite

power system.

Date of issuance: April 29, 1999.

Effective date: April 29, 1999, and shall be implemented in the

next periodic update to the FSAR Update in accordance with 10 CFR

50.71(e).

Amendment Nos.: Unit 1-132; Unit 2-130.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Final Safety Analysis Report Update.

Date of initial notice in Federal Register: October 7, 1998 (63 FR

53952). The supplemental letters dated September 28, 1998, and the

three letters dated February 5, 1999, provided additional clarifying

information, did not expand the scope of the application as originally

noticed, and did not change the staff's original proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 29, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

California

Date of application for amendments: September 3, 1998, as

supplemented by letters dated January 22, 1999, February 5, 1999, and

March 17, 1999.

Brief description of amendments: The amendments change the

Technical Specifications to revise TS 3/4.4.9.1 Figures for heatup and

cooldown to extend their applicability to 16 effective full power

years.

Date of issuance: May 3, 1999.

Effective date: May 3, 1999, to be implemented within 30 days from

the date of issuance.

Amendment Nos.: Unit 1-133; Unit 2-131.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 16, 1998 (63

FR69345). The supplemental letters dated January 22, 1999, February 5,

1999, and March 17, 1999 provided additional clarifying information and

did not change the staff's initial no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Portland General Electric Co

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