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 December 13, 1999, through December 31,

1999. The last biweekly notice was published on December 29, 1999 (64

FR 73083).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

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

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

publication date and page number of this Federal Register notice.

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

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

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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 February 11, 2000, 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 electronically from

the ADAMS Public Library component on the NRC Web site, http://

www.nrc.gov (the Electronic Reading Room). If a request for a hearing

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

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

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

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

the designated Atomic Safety and Licensing Board will issue a notice of

a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

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

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC, and electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,

Unit 1, DeWitt County, Illinois

Date of amendment request: October 25, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specification allowable values for the reactor

protection system electric power monitoring assembly overvoltage and

undervoltage trip setpoints.

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

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consequences of any accident previously evaluated.

The proposed Technical Specification (TS) change revises the

Reactor Protection System (RPS) Electric Power Monitoring Assembly

overvoltage and undervoltage Allowable Values. The new Allowable

Values and setpoints will continue to provide adequate margin to the

normal operating voltage range for the RPS and MSIV [main steam

isolation valve] solenoids, thus minimizing the potential for

inadvertent trips. The proposed change does not have a detrimental

impact on the condition or performance of any plant structure,

system, or component that may initiate an analyzed event. The

proposed change does not physically impact the plant nor does it

impact any design or functional requirements of the associated

system. That is, the proposed change does not degrade the

performance or increase the challenges of any safety systems assumed

to function in the accident analysis. Further, the proposed change

does not impact the Surveillance Requirements themselves nor the way

in which the Surveillances are performed. Consequently, the

probability of an accident previously evaluated is not significantly

increased.

Additionally, the proposed change does not effect the affect the

availability of equipment or systems required for mitigating the

consequences of an accident. The revision of the overvoltage and

undervoltage setpoints will ensure that the associated trip

functions continue to protect the RPS scram solenoids and main steam

isolation valve (MSIV) solenoids so that these devices will perform

their intended safety function. Thus, the affected equipment is

still required to be maintained Operable and capable of performing

the accident mitigation functions assumed in the accident analysis.

As a result, the consequences of any accident previously evaluated

are not significantly affected.

Therefore, based on the above, this change does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed TS change revises the Reactor Protection System

(RPS) Electric Power Monitoring Assembly overvoltage and

undervoltage Allowable Values. The proposed change does not involve

a 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 revised setpoints will continue to

ensure that the RPS bus would be disconnected from its power supply

under specified conditions that could damage the RPS bus powered

equipment. Thus, this change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed change will not involve a significant reduction

in the margin of safety.

The proposed TS change revises the Reactor Protection System

(RPS) Electric Power Monitoring Assembly overvoltage and

undervoltage Allowable Values. The proposed change provides

necessary conservatism in the Allowable Values in the RPS

Surveillance Requirement to ensure that the equipment used to meet

the Limiting Condition for Operation (i.e., each of the two electric

power monitoring assemblies) can continue to perform its required

functions. At the same time, the revised setpoint/Allowable Values

continue to provide adequate margin to the expected operating

voltage range to prevent inadvertent or unnecessary tripping of the

electric power monitoring assemblies (thus preventing unnecessary or

excessive transfer to the alternate power source). The affected

equipment will thus continue to be tested (calibrated and

functionally tested) in a manner that gives confidence that the

equipment can perform its assumed safety function. 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.

Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius

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

NRC Section Chief: Anthony J. Mendiola.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,

Unit 1, DeWitt County, Illinois

Date of amendment request: October 25, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specification definitions for channel

calibrations, channel functional tests, and logic system functional

tests in accordance with Technical Specification Task Force (TSTF)

Standard Technical Specification Change Traveler, TSTF-205, Revision 3,

``Revision of Channel Calibration, Channel Functional Test, and Related

Definitions.''

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 any accident previously

evaluated.

The proposed change clarifies the Technical Specification

requirements for performance of channel calibrations, channel

functional test, and logic system functional tests. Specifically,

the proposed change incorporates the NRC-approved Technical

Specification Task Force (TSTF) Standard Technical Specification

Change Traveler, TSTF-205, Revision 3, ``Revision of Channel

Calibration, Channel Functional Test, and Related Definitions.'' The

change approved per this TSTF is not expected to adversely affect

the performance and effectiveness of required testing as testing

appropriate to the associated Surveillance Requirements will

continue to be performed. The proposed change does not have a

detrimental impact on the condition or performance of any plant

structure, system, or component that initiates an analyzed event.

Consequently, the probability of an accident previously evaluated is

not significantly increased. The equipment being tested is still

required to be operable and capable of performing the accident

mitigation functions assumed in the accident analysis. As a result,

the consequences of any accident previously evaluated are not

significantly affected. Therefore, this change does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The scope of the proposed change is limited to the clarification

of existing test requirements. As such, the proposed change does not

involve a 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. Thus, this change does not create the

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

previously evaluated.

3. The proposed change will not involve a significant reduction

in the margin of safety.

As noted above, the proposed change clarifies requirements for

the performance of channel calibrations, channel functional tests,

and logic system functional tests. Specifically, the proposed change

incorporates the NRC-approved Technical Specification Task Force

(TSTF) Standard Technical Specification Change Traveler, TSTF-205,

Revision 3, ``Revision of Channel Calibration, Channel Functional

Test, and Related Definitions.'' No changes or setpoints to plant

process limits are involved. The surveillance requirements as

revised will continue to ensure that affected equipment is tested in

a manner that gives confidence that the equipment can perform its

appropriate safety function. 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.

Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius

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

NRC Section Chief: Anthony J. Mendiola.

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AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,

Unit 1, DeWitt County, Illinois

Date of amendment request: December 16, 1999.

Description of amendment request: The proposed amendment would

allow a one-time extension of some Technical Specification surveillance

intervals to support elimination of a planned spring 2000 mid-cycle

outage (PO-8). For the applicable surveillances, the licensee proposes

to extend their current surveillance intervals to November 30, 2000,

the scheduled startup date from refueling outage 7 (RF-7).

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 any accident previously

evaluated.

The proposed Technical Specification (TS) changes involve a one-

time only change in the surveillance test intervals of selected

Surveillance Requirements (SRs). The proposed TS changes do not

impact the TS surveillance performance requirements themselves nor

the way in which the surveillances are performed. The proposed TS

changes do not physically involve any changes to the plant, nor do

they impact any design or functional requirements of the associated

systems. Thus, the proposed TS changes do not increase the

challenges of any safety systems assumed to function in the accident

analysis.

In addition, the proposed TS changes do not significantly affect

the availability of equipment or systems required to mitigate the

consequences of an accident because (1) extension of the test

intervals to the extent requested is not expected to have a

significant impact on availability (i.e., no extended test interval

would exceed 30 months), and (2) other or more frequent testing

performed for the affected systems or components, as well as for

redundant systems or components, supports continued availability of

the affected function. The equipment subject to testing per the

affected SRs is still required to be operable and capable of

performing any accident mitigation functions assumed in the accident

analysis. Furthermore, a historical review of surveillance test

results identified no failures that would invalidate these

conclusions.

Based on the above, the proposed TS changes do not significantly

increase the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed TS changes involve a one-time only change in the

surveillance testing intervals of selected SRs. Such changes do not

introduce any failure mechanisms of a different type than those

previously evaluated since there are no physical changes being made

to the facility. In addition, the surveillance test requirements

themselves, and the way surveillance tests are performed, will

remain unchanged. Therefore, the proposed TS changes do not create

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The one-time extended surveillance frequencies do not result in

a significant reduction in the margin of safety. Although the

proposed TS changes will result in an increase in the interval

between surveillance tests, the impact, if any, on system

availability is small. This is because, as noted previously,

extension of the test intervals to the limited extent proposed would

not be expected to have a significant impact on availability. Other

or more frequent testing performed for the affected systems or

components, as well as the testing performed for redundant systems

or components, supports continued availability of the affected

functions.

In addition, the proposed changes do not involve any physical

changes to the affected systems or components, nor do they involve

any changes to setpoints, operating limits, or safety limits.

Based on the above, the assumptions in the licensing basis are

not impacted, and the proposed TS changes do not significantly

reduce a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius

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

NRC Section Chief: Anthony J. Mendiola.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas

Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas

Date of amendment request: September 17, 1999.

Description of amendment request: The proposed change to the

Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Technical

Specifications would lower the maximum limit for contents of the

gaseous radioactive system from 300,000 curies (Ci) to 78,782 Ci and

82,400 Ci for ANO-1 and ANO-2, respectively. This limit would ensure

that, upon an uncontrolled release of the tank's contents over a 2-hour

period, the resulting total whole body exposure to a member of the

public at the nearest exclusion area boundary would not exceed 0.5

roentgen equivalent man (rem).

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:

An evaluation of the proposed change has been performed in

accordance with 10CFR50.91(a)(1) regarding no significant hazards

considerations using the standards in 10CFR50.92(c). A discussion of

these standards as they relate to this amendment request follows:

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

or Consequences of an Accident Previously Evaluated

The proposed change to lower the current technical specification

(TS) gas storage tank activity limits does not require new hardware

or physical equipment modifications to the plant design. By lowering

the setpoint, the resultant exposure at the exclusion area boundary

upon an inadvertent release of a gas storage tank's content will be

limited to 0.5 rem. Therefore the consequences of such an

uncontrolled release of activity are effectively reduced.

Additionally, no new accident is introduced by the proposed

reduction in activity limits associated with the gas storage tanks.

Therefore, reducing the gas storage tank limits from 300,000

Curies (Ci) to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2,

respectively) 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 change affects the consequences of an event

associated with the loss of gas storage tank radioactive contents on

either ANO-1 or ANO-2. Since this event has been previously

evaluated, no new or different accident can be associated with the

proposed change. Decreasing the present TS activity limits results

in an exposure at the exclusion area boundary to be limited to 0.5

rem in the event of an inadvertent release of a gas storage tank's

content.

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

or different kind of accident from any previously evaluated.

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

Safety

The proposed change conservatively lowers the existing TS GRW

[Gaseous Radwaste] System gas storage tank activity limits from

300,000 Ci to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2

respectively). In doing so, the resultant exposure to a member of

the public at the exclusion area boundary during an inadvertent

release of gas storage tank contents over a two-hour period is

reduced to 0.5 rem or less. The proposed change, therefore, retains

the margin of safety for both ANO-1 and ANO-2.

Therefore, this change does not involve a significant reduction

in the margin of safety.

Therefore, based on the reasoning presented above and the

previous discussion

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of the amendment request, Entergy Operations, Inc. has determined

that the requested change does not involve a significant hazards

consideration.

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.

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

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

NRC Section Chief: Robert A. Gramm.

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, 1999.

Description of amendment request: The proposed license amendment

request would revise Fuel Handling Accident (FHA) dose calculations for

3 scenarios documented in the River Bend Station, Unit 1 (RBS), Updated

Safety Analysis Report (USAR). The first is a FHA in the fuel building,

assumed to occur 24 hours post-shutdown. A second FHA analysis was

prepared to support Amendment 35 to RBS Technical Specifications (TS)

which assumed a FHA occurs in the primary containment 80 hours post-

shutdown during Local Leakage Rate Testing (LLRT). A third analysis was

prepared in support of Amendment 85 to the RBS TS which assumed the

containment is open at 11 days.

These analyses are being updated to account for several changes.

The primary reason for the revisions, as stated by the licensee, was to

update the analyses to reflect current RBS operating strategies and

make the analyses consistent with each other. Specifically, Cases 1 and

2 of the three analyses assumed a Radial Peaking Factor (RPF) of 1.5

consistent with Regulatory Guide (RG) 1.25. However, current core

design strategies could lead to an RPF as high as 1.65. In addition, to

account for the potential impact of extended burnup fuel in future

operating cycles, an increased iodine-131 gap fraction of 0.12 was more

conservatively assumed in lieu of the 0.10 recommended by RG 1.25. The

revised analysis also includes a change to the control room atmospheric

dispersion factors (/Q) for the Main Control Room (MCR)

ventilation system. Credit is taken for Standard Review Plan (SRP)

Section 6.4 guidance for manual dual control room air intakes in that

the /Q's are divided by 4. The revised FHA analyses also

credit this action at a 20 minute delay to be consistent with the Loss

of Coolant Accident (LOCA) analysis.

Furthermore, an error was discovered in one of the FHA

calculations. The release rate assumed in the analysis did not ensure

that the RG 1.25 assumption of a 2-hour release was preserved. The

error is the result of an inherent bias in the secondary mixing effects

in the dose calculation. The results continue to be bounded by the

guidance contained in SRP 15.7.4 and RG 1.25.

Reanalysis showed that the release rate error, compounded with the

other changes discussed above, resulted in calculated doses greater

than those currently found in the RBS USAR. In addition, some of the

doses were also greater than those presented in the Amendment 85

submittal. However, the licensee has stated that the results of the

revised analyses remain ``well within'' 10 CFR 100, the guidance

contained in SRP 15.7.4, and RG 1.25. Since the analyses results are

above those reported in the RBS USAR, the criterion of 10 CFR

50.59(a)(2)(i) is, therefore, satisfied. Accordingly, the licensee has

concluded that these changes involve an unreviewed safety question.

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 significantly increase the

probability or consequences of an accident previously evaluated.

The analyses changes described by this proposed change to the

USAR are not initiators to events, and, therefore, do not involve

the probability of an accident. The changes to the FHA calculations

for radiological doses following a FHA reflect the current operating

strategies and make the analyses consistent. These changes included:

Accounting for the impact of extended burnup fuel,

Addressing a change to the control room atmospheric

dispersion factors assumed in the analysis, and

Revising the Radial Peaking Factor (RPF) used in the

analysis. Current core design strategies could lead to a RPF higher

[than] that assumed in Regulatory Guide 1.25.

The TRANSACT code is used for offsite dose and control room dose

calculations. The TRANSACT code is derived from the TACT V code

documented in

NUREG/CR-5109. RBS has benchmarked the TRANSACT code as

discussed in the request dated August 17, 1995, (RBG-41728) which

resulted in the NRC granting Amendment 85.

The revisions to the FHA are used to establish operational

conditions where specific activities represent situations where

significant radioactive releases can be postulated. These

operational conditions include:

Initial fuel movement in the Fuel Building 24 hours

after shutdown,

Fuel movement in Primary Containment after 80 hours

with leakrate testing being conducted, and

Fuel movement in Primary Containment with the Primary

Containment open.

Because the analyses affected by the changes are not considered

an initiator to any previously analyzed accident, these changes

cannot increase the probability of any previously evaluated

accident. Therefore, this change does not increase the probability

of occurrence of an accident evaluated previously in the safety

analysis report (SAR).

This proposed change to the USAR does increase the consequences

of an accident, but the increase is within all regulatory limits and

guidance. While the calculated off-site and control room doses of a

FHA did increase, the dose consequences remain below the regulatory

limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC [General Design

Criterion]-19 as approved per NUREG-0989, and the guidance contained

in SRP 15.7.4 of less than 25% of the 10 CFR 100 limits. The cause

of these events remains the failure of the fuel assembly lifting

mechanism. These analyses demonstrated that for the worst case

bundle drop, the regulatory dose guidelines of SRP 15.7.4 continue

to be satisfied for the required decay periods.

This change accounts for the potential effects of current fuel

design and operating strategies including increased burnup of fuel,

increased iodine-131 fraction released, Main Control Room

ventilation system operation, and release rate timing assumptions.

Reanalysis of the off-site dose calculation demonstrates that the

revised doses are increased but remain less than the regulatory

limits of 10 CFR 100 and within the guidance of SRP 15.7.4.

Therefore, this change does not significantly increase the

consequences of an accident previously evaluated in the SAR.

The proposed changes, in conjunction with existing

administrative controls, bound the conditions of the current design

basis fuel handling accident analysis. The analysis also concludes

the limiting offsite radiological consequences are well within the

acceptance criteria of NUREG[-]0800, Section 15.7.4 and 10 CFR 50,

Appendix A, GDC[-]19. The analysis is also conducted in a

conservative manner containing margins in the calculation of

mechanical analysis, iodine inventory, and iodine decontamination

factor. Each of these conservatisms will further decrease the

consequences. Therefore, the proposed changes do not significantly

increase the probability or consequences of any previously evaluated

accident.

2. The proposed changes would not create the possibility of a

new or different kind of accident from any previous[ly] analyzed.

This change does not involve initiators to any events in the

SAR, nor does the activity create the possibility for any new

accidents. Rather, this change is a result of the

[[Page 1923]]

evaluation of the most limiting FHA, which can occur at River Bend.

The proposed changes to the dose analyses are consistent with

previous limits, only revising previous evaluations to account for

current operating strategies and assumptions. These changes

included:

Accounting for the impact of extended burnup fuel,

addressing a change to the control room atmospheric

dispersion factors assumed in the analysis, and

Revising the Radial Peaking Factor (RPF) used in the

analysis. Current core design strategies could lead to a RPF higher

[than] that assumed in Regulatory Guide 1.25.

The radiological consequences remain within accepted limits of

10 CFR 100 and guidance of the Standard Review Plan (NUREG-0800)

Section 15.7.4. Therefore, these changes are consistent with the

design basis analysis. The proposed changes do not introduce any new

modes of plant operation and do not involve physical modifications

to the plant. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any

previous[ly] analyzed.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The dose consequences are calculated in accordance with

regulatory guidance found in Regulatory Guide 1.25 and the SRP

[S]ection 15.7.4. The RBS analyses conservatively assumed that

failures are consistent with those in the standard General Electric

GESTAR II. These analyses result in a bounding number of fuel

failures. The RBS analyses are also consistent with those approved

by the NRC [Nuclear Regulatory Commission] in support of Technical

Specification Amendments 35 and 85 to the River Bend Station license

(NPF-47). The radiological dose consequences resulting from these

failures are therefore analyzed using accepted methods and criteria.

In addition, the analyses contain known conservatisms and margins to

ensure the results will remain bounding.

The revised limits are used to establish operational conditions

where specific activities represent situations where significant

radioactive releases can be postulated. These operational conditions

are consistent with the design basis analysis and are established

such that the radiological consequences are at or below the current

regulatory limits and guidance. Safety margins and analytical

conservatisms have been evaluated and are well understood.

Conservative methods of analysis are maintained through the use of

accepted methodology and benchmarking the proposed methods to

previous analysis. Margins are retained to ensure that the analysis

adequately bounds all postulated event scenarios. The proposed

change only eliminates some excess conservatism from the analysis.

In addition, EOI [Entergy Operations, Inc.] has implemented

NUMARC [Nuclear Management and Resources Council (now NEI)] 91-06

guidelines for shutdown operations at RBS. Shutdown Operations

Protection Plan and Primary-Secondary Containment Integrity

procedures presently include guidance for closure of the containment

hatch and other significant openings in containment, in addition to

the requirements contained in the license and design basis. This

additional protection will enhance the ability to limit offsite

effects.

Acceptance limits for the fuel handling accident are provided in

10 CFR 100 with additional guidance provided in NUREG[-]

0800, Section 15.7.4. The proposed changes continue to ensure that the

whole-body and thyroid doses at the exclusion area and low population

zone boundaries, as well as control room doses, are below the

corresponding regulatory limits. These margins are unchanged,

therefore, the proposed changes do not involve a significant reduction

in a margin of safety.

The commission has provided guidance concerning the application

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

FR 7751, March 6, 1986) of amendments that are not considered likely

to involve a significant hazards consideration. This proposed

amendment is very similar to example (vi):

(vi) A change which either may result in some increase to the

probability or consequences of a previously-analyzed accident or may

reduce in some way a safety margin, but where the results of the

change are clearly within all acceptable criteria with respect to

the system or component specified in the Standard Review Plan: for

example, a change resulting from the application of a small

refinement of a previously used calculational model or design

method.

As we have shown in the preceding discussion, this refinement to

the FHA dose calculation results in a small increase to the

consequences of a previously analyzed accident, but the results of

the change remain clearly within the guidelines of 10 CFR 100,

Appendix A, GDC[-]19, and the guidance of SRP [S]ection 15.7.4,

without reducing 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.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

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

NRC Section Chief: Robert A. Gramm.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: November 17, 1999 (L-99-241)

Description of amendment request: The proposed amendment would

revise the St. Lucie Unit 1 and Unit 2 Technical Specifications (TS) to

require laboratory testing of activated charcoal samples for applicable

engineered safety feature ventilation systems using the American

Society for Testing and Materials (ASTM) D3803-1989 protocol. The

affected TS are Units 1 and 2 shield building ventilation system, TS

4.6.6.1; Unit 1 emergency core cooling system area ventilation system,

TS 4.7.8.1; Unit 1 control room emergency ventilation system, TS

4.7.7.1; Unit 2 control room emergency air cleanup system, TS 4.7.7;

and Unit 1 fuel pool ventilation system--fuel storage, TS 4.9.12.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of any accident previously

evaluated. The new charcoal testing protocol is performed offsite on

samples extracted from the safety-related ventilation systems.

Therefore, there is no impact on any accident initiator and

therefore, no changes on the probability. The proposed testing

protocol is more conservative than previous tests; therefore, the

efficiency of charcoal for the affected safety-related systems would

not be overestimated. With the new testing protocol, more

conservative testing results are expected since the temperature at

which testing is performed is lower and the charcoal retention

capability is more consistent with actual accident conditions. The

proposed change thus ensures that the charcoal in service will

comply with the penetration requirements to meet the design basis

accident conditions.

Therefore, operation of the facility in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously

evaluated. The proposed new charcoal testing protocol only affects

surveillance testing requirements for ventilation systems. The

functions of these systems remain unchanged and unaffected. No new

system interactions have been introduced by the proposed amendment,

which would create a new or different type of accident than

previously analyzed. No physical changes are being made to any

structure, system or component. The operation of the facility has

not been altered by the proposed amendment. The systems involved are

not considered to initiate any accidents as previously evaluated.

[[Page 1924]]

The proposed amendment will not change the physical plant or the

modes of operation defined in the facility license. The changes do

not involve the addition of new equipment or the modification of

existing equipment, nor do they alter the design of St. Lucie plant

systems. Therefore, operation of the facility in accordance with the

proposed amendment would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment does not involve a reduction in the

margin of safety. The margin of safety of the Technical

Specifications, its bases, the Final Safety Analysis Report, the

Safety Evaluation Report or in any other design document has not

been affected by the proposed amendment. The change provided in this

proposed amendment is related to introducing an improved testing

protocol for the activated charcoal in safety related ventilation

systems. The change consists of testing the charcoal with a new

testing protocol and with lower test temperatures to resemble more

closely accident conditions and to eliminate potential

overestimation of charcoal efficiency.

Therefore, operation of the facility in accordance with the

proposed amendment would 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.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420

NRC Section Chief: Richard P. Correia

GPU Nuclear, Inc., et al., Three Mile Island Nuclear Station, Unit 2

(TMI-2), Docket No. 50-320, Dauphin County, Pennsylvania

Date of amendment request: November 5, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) Sections 6.8.1.4, 6.5.4.6, 6.13,

6.14, and 6.8.3. Specifically, Sections 6.13, 6.14 and 6.8.3 would be

revised to eliminate the requirement to notify the Nuclear Regulatory

Commission (NRC) of exceeding environmental limits and changes to

environmental permits such as National Pollution Discharge Elimination

System (NPDES). The requirements contained in the individual

environmental permits and program regulations administered by the U.S.

Environment Protection Agency (EPA), Pennsylvania Department of

Environmental Protection (PADEP), and other regulatory agencies with

program jurisdiction for reporting are included in plant procedures and

data base tracking systems. Sections 6.8.1.4 and 6.5.4.6 are changes to

the amendment that are administrative in nature and reflect a

streamlining of the GPU Nuclear, Inc. management structure.

Basis for proposed no significant hazards consideration

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

provided its analyses of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes to the TMI-2 [Three Mile Island, Unit 2]

Technical Specifications do not involve a significant increase in

the probability of occurrence or consequences of an accident or

malfunction of equipment important to safety previously analyzed in

the safety analysis report. The changes have no impact on plant

operations or the release of radioactive materials.

2. The proposed changes to the TMI-2 Technical Specifications

will not create the possibility for an accident or malfunction of a

different type than any previously evaluated in the safety analysis

report because no plant configuration or operational changes are

involved.

3. The changes will not involve a significant reduction in the

margin of safety as defined in the basis for any technical

specification for TMI-2 because no change to operational limits will

be made.

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.

Attorney for Licensee: Ernest L. Blake, Jr Esq., Shaw, Pittman,

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

NRC Section Chief: Mike Masnik.

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

County, Iowa

Date of amendment request: November 10, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 5.5.7.c, to commit to the American

Society for Testing and Materials (ASTM) D3803-1989 test protocol for

the ventilation filter testing program. The proposed changes are

consistent with Attachment 2, Sample Technical Specifications, in

Generic Letter 99-02. Because the current TS penetration limits do not

reflect a safety factor in excess of that assumed in the dose

calculations of the accident analysis, the TS change request would also

revise the allowable penetration values to correspond to a safety

factor of 2.

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 for the administrative changes:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The ESF [engineered safety feature] ventilation systems are not

initiators of any accident previously evaluated and the change in

testing protocol to ASTM D3803-1989 as requested by the NRC will be

more accurate and realistic and provide greater assurance of

consistency. The acceptance criteria will be more conservative than

those currently used in TS 5.5.7.c.

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

new or different kind of accident from any accident previously

evaluated.

No new types of accidents are being introduced because no

modifications or changes in operations are being proposed for the

ESF [engineered safety feature] ventilation systems. The proposed

changes to TS 5.5.7.c impact acceptance criteria and test protocols

only.

3. The proposed amendment will not involve a significant

reduction in a margin of safety.

The margin of safety is not reduced. The proposed change in ESF

ventilation testing protocol includes a safety factor of two (2) for

the penetration limit in excess of that assumed in the dose

calculations of the DAEC [Duane Arnold Energy Center] accident

analysis.

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.

Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Claudia M. Craig.

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

County, Iowa

Date of amendment request: November 22, 1999.

Description of amendment request: The proposed amendment would

adopt selected NRC-approved generic changes to the Improved Technical

Specifications (ITS) NUREGs. The 16 changes come from the Technical

Specification Task Force (TSTF) process

[[Page 1925]]

developed by the industry and the NRC. Three of these changes are

Bases-only changes but are included for completeness relative to the

TSTF process.

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 for the administrative changes:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and

rewording the existing Technical Specifications. The reformatting,

renumbering, and rewording process involves no technical changes to

the existing Technical Specifications. As such, this change is

administrative in nature and does not affect initiators of analyzed

events or assumed mitigation of accident or transient events.

Therefore, this change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of

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

or changes in methods governing normal plant operation. The proposed

change will not impose any new or eliminate any old requirements.

Thus, this change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

of safety?

The proposed change will not reduce a margin of safety because

it has no effect on any safety analyses assumptions. This change is

administrative in nature. Therefore, the change does not involve a

significant reduction in a margin of safety.

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 for more restrictive changes:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for

operation of the facility. These more stringent requirements do not

result in operation that will increase the probability of initiating

an analyzed event and do not alter assumptions relative to

mitigation of an accident or transient event. The more restrictive

requirements continue to ensure process variables, structures,

systems, and components are maintained consistent with the safety

analyses and licensing basis. Therefore, this change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of

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

or changes in methods governing normal plant operation. The proposed

change does impose different requirements. However, these changes

are consistent with the assumptions in the safety analyses and

licensing basis. Thus, this change does not create the possibility

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

evaluated.

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

of safety?

The imposition of more restrictive requirements either has no

effect on or increases the margin of plant safety. As provided in

the justification, each change in this category is, by definition,

providing additional restrictions to enhance plant safety. The

change maintains requirements within the safety analyses and

licensing basis. Therefore, the change does not involve a

significant reduction in a margin of safety.

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 for less restrictive changes--removed detail:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relocates certain details from the Technical

Specifications to other documents under regulatory control. The

Bases, UFSAR [updated final safety analysis report], and Technical

Requirements Manual will be maintained in accordance with 10 CFR

50.59. In addition to 10 CFR 50.59 provisions, the Technical

Specification Bases are subject to the change control provisions in

the Administrative Controls Chapter of the Technical Specification.

The UFSAR is subject to the change control provisions of 10 CFR

50.71(e). Other documents are subject to controls imposed by

Technical Specifications or regulations. Since any changes to these

documents will be evaluated, no significant increase in the

probability or consequences of an accident previously evaluated will

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

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a 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. Does this change involve a significant reduction in a margin

of safety?

The proposed change will not reduce a margin of safety because

it has no effect on any safety analyses assumptions. In addition,

the details to be moved from the Technical Specifications to other

documents are the same as the existing Technical Specifications.

Since any future changes to these details will be evaluated, no

significant reduction in a margin of safety will be allowed. A

significant reduction in the margin of safety is not associated with

the elimination of the 10 CFR 50.92 requirement for NRC review and

approval of future changes to the relocated details. The proposed

change is consistent with the BWR [Boiling Water Reactor]/4 Standard

Technical Specifications, NUREG-1433, issued by the NRC Staff,

revising the Technical Specifications to reflect the approved level

of detail, which indicates that there is no significant reduction in

the margin of safety.

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 for less restrictive changes--category 3, relaxation of

completion time:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relaxes the Completion Time for a Required

Action. Required Actions and their associated Completion Times are

not initiating conditions for any accident previously evaluated and

the accident analyses do not assume that required equipment is out

of service prior to the analyzed event. Consequently, the relaxed

Completion Time does not significantly increase the probability of

any accident previously evaluated. The consequences of an analyzed

accident during the relaxed Completion Time are the same as the

consequences during the existing Completion Time. As a result, the

consequences of any accident previously evaluated are not

significantly increased. Therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a 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

Required Actions and associated Completion Times have been evaluated

to ensure that no new accident initiators are introduced. Thus, this

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

accident from any accident previously evaluated.

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

of safety?

The relaxed Completion Time for a Required Action does not

involve a significant reduction in the margin of safety. As provided

in the justification, the change has been evaluated to ensure that

the allowed Completion Time is consistent with the safe operation

under the specified Condition, considering the operability status of

the

[[Page 1926]]

redundant systems of required features, the capacity and capability

of remaining features, a reasonable time for repairs or replacement

of required features, and the low probability of a DBA [design basis

accident] occurring during the repair period. Therefore, this change

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

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 for less restrictive changes--category 4, relaxation of

required action.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relaxes Required Actions. Required Actions

and their associated Completion Times are not initiating conditions

for any accident previously evaluated and the accident analyses do

not assume that required equipment is out of service prior to the

analyzed event. Consequently, the relaxed Required Actions do not

significantly increase the probability of any accident previously

evaluated. The Required Actions in the change have been developed to

provide assurance that appropriate remedial actions are taken in

response to the degraded condition, considering the operability

status of the redundant systems of required features, and the

capacity and capability of remaining features while minimizing the

risk associated with continued operation. As a result, the

consequences of any accident previously evaluated are not

significantly increased. Therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a 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

Required Actions and associated Completion Times in the change have

been evaluated to ensure that no new accident initiators are

introduced. Thus, this change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

of safety?

The relaxed Required Actions do not involve a significant

reduction in the margin of safety. As provided in the justification,

the change has been evaluated to minimize the risk of continued

operation under the specified Condition, considering the operability

status of the redundant systems of required features, the capacity

and capability of remaining features, a reasonable time for repairs

or replacement of required features, and the low probability of a

DBA [design basis accident] occurring during the repair period.

Therefore, this change does not involve a significant reduction in a

margin of safety.

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 for less restrictive changes--category 6, relaxation of

surveillance requirement acceptance criteria:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relaxes the acceptance criteria of

Surveillance Requirements. Surveillances are not initiators to any

accident previously evaluated. Consequently, the probability of an

accident previously evaluated is not significantly increased. The

equipment being tested is still required to be Operable and capable

of performing the accident mitigation functions assumed in the

accident analysis. As a result, the consequences of any accident

previously evaluated are not significantly affected. Therefore, this

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a 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. Thus,

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

kind of accident from any accident previously evaluated.

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

of safety?

The relaxed acceptance criteria for Surveillance Requirements do

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

provided in the justification, the relaxed Surveillance Requirement

acceptance criteria have been evaluated to ensure that they are

sufficient to verify that the equipment used to meet the LCO

[limiting condition for operation] can perform its required

functions. Thus, appropriate equipment continues to be tested in a

manner that gives confidence that the equipment can perform its

assumed safety function. 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.

Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Claudia M. Craig.

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

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

California

Date of amendment requests: September 3, 1998, as supplemented by

letters dated January 22, February 5, March 17, and November 24, 1999.

The September 3, 1998, amendment application was previously noticed in

the Federal Register on December 16, 1998 (63 FR 69345).

Description of amendment requests: The amendment would revise

Section 5.6.6, ``Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE

LIMITS REPORT (PTLR),'' of the improved Technical Specifications (TSs),

that were issued in Amendment Nos. 135 and 135 on May 28, 1999. The

amendment would add the phrase ``and LTOP'' (low-temperature

overpressure protection) to the first sentence of item 5.6.6.b that

identifies the limits that can be determined by the licensee in the

PTLR, and (2) replace the current list of documents listed in item

5.6.6.b by the NRC letter that would approve this amendment and

Westinghouse WCAP-14040-NP-A, ``Methodology Used to Develop Cold

Overpressure Mitigation System Setpoints and RCS Heatup and Cooldown

Limit Curves,'' dated January 1996. WCAP-14040-NP-A is the NRC-approved

topical report which provides a methodology for developing the LTOP

setpoints and RCS heatup and cooldown limit curves for Westinghouse

plants, such as Diablo Canyon Nuclear Power Plant, Units 1 and 2.

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 changes to Figures 3.4-2 and 3.4-3 of Technical

Specification (TS) 3.4.9.1 and the associated Bases adjust the

reactor coolant system (RCS) heatup and cooldown pressure/

temperature (P/T) limits to permit operation through 16 effective

full power years (EFPY). The 16 EFPY P/T limits are more restrictive

than the current limits; this accounts for an expected incremental

increase in reactor vessel embrittlement, and assures the reactors

will continue to be operated within acceptable stresses and at

temperatures for which the reactor vessel metal exhibits ductile

properties. The P/T limits developed for 16 EFPY were determined in

accordance with 10 CFR 50, Appendix G, and maintain the same margins

of safety as the current limits. The proposed changes will not

impact the probability of overpressurization or brittle fracture of

the vessel, and therefore will not impact the consequences of an

accident.

The present low temperature overpressure protection (LTOP)

pressure and enable temperature setpoints were reviewed and found to

be acceptable and conservative for

[[Page 1927]]

use through 16 EFPY, based on use of ASME [American Society of

Mechanical Engineers] Code Case N-514, which provides acceptable

margins to the prevention of vessel overpressurization and brittle

fracture. Therefore, there is no change to the consequences of

accidents previously analyzed. Since no changes are proposed in the

actual LTOP setpoints, nor any physical alteration of the LTOP

system, nor a change to the method by which the LTOP system performs

its function, there would be no change to the probability of an

accident previously evaluated. The proposed change to the Bases

incorporates use of ASME Code Case N-514, which will benefit DCPP

[Diablo Canyon Power Plant] by not resulting in a reduced RCS P/T

window and reduced power-operated relief valve (PORV) pressure

setpoint for LTOP. This maintains the current level of operator

flexibility during heatup and cooldown, and prevents an increase in

the probability of an accident associated with an inadvertent PORV

actuation.

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 to TS 3.4.9.1, ``Reactor Coolant System--

Pressure/Temperature Limits,'' do not involve any physical

alteration to any plant system or change the method by which any

safety-related system performs its function. The changes to TS

3.4.9.1 account for the effects of an incremental increase in

reactor vessel embrittlement and are requested in order to restrict

future reactor operation to within acceptable stress levels and

temperature regimes in accordance with 10 CFR 50, Appendix G,

requirements. These changes are needed to maintain the current P/T

limit margins of safety as defined by 10 CFR 50, Appendix G, and

ASME XI, Appendix G, for operation through 16 EFPY. The possibility

of a new kind of accident such as catastrophic failure of the

reactor vessel is prevented by maintaining acceptable margins of

safety.

The present LTOP pressure setpoint was reviewed and found to be

acceptable and conservative for the extension of the P/T curves to

16 EFPY.

Additionally, the proposed changes will not affect the ability

of the LTOP system to provide pressure relief at low temperatures,

thereby maintaining the LTOP design basis.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes to TS 3.4.9.1 adjust the RCS heatup and

cooldown P/T limits to permit operation through 16 EFPY. The P/T

limits have been determined in accordance with 10 CFR 50, Appendix

G, and include the safety margins with regard to brittle fracture

required by the ASME Section XI, Appendix G, which maintain the same

margins of safety as the current limits.

The LTOP setpoints were reevaluated using the requirements of

ASME Code Case N-514. This code case was developed to provide the

necessary margins of safety for the prevention of reactor vessel

overpressurization and brittle fracture. The LTOP evaluation results

conclude the current LTOP setpoints are conservative for operation

through 16 EFPY. In addition, avoiding an unnecessary reduction in

the LTOP, the PORV pressure setpoint prevents an increase in the

likelihood of an inadvertent PORV actuation

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

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

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

NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power

Plant, Unit 3, Humboldt County, California

Date of amendment request: December 1, 1999.

Description of amendment request: The proposed amendment would

revise the Humboldt Bay Power Plant (HBPP) Unit 3 Technical

Specifications (TS) related to fire protection, administrative

controls, and quality assurance audits. The fire protection

requirements would be relocated verbatim from the TS to the HBPP

Defueled Safety Analysis Report (DSAR). The administrative controls

requirements would be revised to (1) refer to the DSAR for a

description of the plant organization, (2) modify information

pertaining to plant staff titles and qualifications to reflect the

current organization, and (3) replace a reference to the Final Hazards

Summary Report with a reference to the DSAR. Quality assurance audit

requirements would be relocated from the TS to the Quality Assurance

Plan.

Basis for proposed no significant hazards consideration

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

provided its analyses of the issue of no significant hazards

consideration, which are presented below:

For the proposed changes to the fire protection requirement, the

licensee's analysis states:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The FPP [Fire Protection Program] and FPS [Fire Protection

System] are not being changed. Operability requirements and

procedural controls of the FPP and FPS are not being changed. The

proposed changes involve only where the FPP and FPS description is

located and how changes can be made. Consequently, the changes will

not affect the probability or consequences of an accident occurring.

Future changes to the FPP and FPS as described in the Defueled

Safety Analysis Report would be made in accordance with 10 CFR

50.59. This ensures that adequate controls will remain in place so

that the public health and safety will be protected.

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 FPP and FPS are not being changed. Operability requirements

and procedural controls of the FPP and FPS are not being changed.

The proposed changes involve only where the FPP and FPS description

is located and how changes can be made. Consequently, the changes

will not affect the probability or consequences of an accident

occurring.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The FPP and FPS are not being changed. Operability requirements

and procedural controls of the FPP and FPS are not being changed.

The proposed changes involve only where the FPP and FPS description

is located and how changes can be made. Consequently, the changes

will not affect the probability or consequences of an accident

occurring.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

For the proposed changes to the administrative controls

requirements, the licensee's analysis states:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The organization title and responsibility changes update the

Technical Specification (TS) to reflect the current organization and

have no impact on the function or operability of plant systems,

structures, or components, or the ability of the plant to safely

maintain SAFSTOR status. Consequently, the changes will not affect

the probability or consequences of an accident occurring.

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

[[Page 1928]]

accident from any accident previously evaluated.

The organization title and responsibility changes update the TS

to reflect the current organization and have no impact on the

function or operability of plant systems, structures, or components,

or the ability of the plant to safely maintain SAFSTOR status.

Consequently, the changes will not affect the probability or

consequences of an accident occurring.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The organization title and responsibility changes update the TS

to reflect the current organization and have no impact on the

function or operability of plant systems, structures, or components,

or the ability of the plant to safely maintain SAFSTOR status.

Consequently, the changes will not affect the probability or

consequences of an accident occurring.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

For the proposed changes to the quality assurance audit

requirements, the licensee's analysis states:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes simplify the Technical Specifications (TS),

meet regulatory requirements for relocated TS, and implement: (1)

The recommendations of NRC's letter dated October 25, 1993, from

William T. Russell to the chairpersons of the industry owners

groups; (2) the Commission's Final Policy Statement on TS

Improvements; and (3) the current revision of 10 CFR 50.36. Future

changes to these requirements will be controlled by 10 CFR 50.54.

This ensures that adequate controls will remain in place so that the

public health and safety will be protected. The proposed changes are

administrative in nature and do not involve any modifications to any

plant equipment or affect plant operation.

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 are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety-related system performs

its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not alter implementation of the basic

regulatory requirements and do not affect any safety analyses.

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 requests involve no significant hazards consideration.

Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas

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

NRC Section Chief: Michael Masnik.

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

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

Date of application for amendments: October 12, 1999 (TS 99-15).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79

(Unit 2) by revising the Technical Specification (TS) to provide for

unisolation of containment penetrations under administrative controls.

This revision will add a footnote to Specification 3.9.4.c indicating

this allowance and the necessary Bases addition for this section to

clarify the use of this allowance.

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 will allow the opening of specific

containment penetrations during the movement of irradiated fuel or

core alterations provided administrative controls are implemented.

These controls will establish the proper awareness of the unisolated

penetration condition, designate individuals to isolate the

penetration in the event of an FHA [fuel handling accident], and

[to] ensure the auxiliary building gas treatment system (ABGTS) is

available. The status of containment penetrations does not impact

the generation of an accident nor does the ability to unisolate

penetrations affect this potential. The proposed revision does not

alter any plant equipment or operating practices other than

penetration isolation such that the probability of an accident is

increased.

The administrative controls provide adequate requirements to

provide timely identification and closure of penetrations opened

under this allowance should a fuel handling event occur. Designated

individuals ensure that adequate resources are available to isolate

the penetration such that the offsite dose consequences are not

significantly impacted. The lack of motive force in containment

during fuel movement to expel the radioactive material allows a more

flexible isolation interval. The exception for the containment

ventilation isolation valves is based on being exposed to a motive

force and the flow paths outside the auxiliary building secondary

containment enclosure (ABSCE) is based on being exposed to an

unfiltered atmosphere. Timely isolation of the specified flow paths

is required to ensure that the unlikely transmission of radioactive

material does not occur. Interactions that may occur during the

period of time before isolation will be controlled by operation of

the ABGTS and will not significantly increase the consequences of an

accident as previously evaluated. Completion of penetration

isolation and operation of the ABGTS, as required by the

administrative controls, will maintain the offsite dose consequences

well within the 10 CFR 100 limits.

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

new or different kind of accident from any accident previously

evaluated.

The proposed allowance to open penetrations in Mode 6 will not

alter plant functions or equipment operating practices other than

penetration isolation. Containment penetration status is not

considered to be the source of an accident. Therefore, since the

plant functions and equipment are not altered and the isolation

status of containment penetrations do not contribute to the

initiation of postulated accidents, the proposed revision will not

create a new or different kind of accident.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The isolation requirements for containment penetrations ensure

that the release of radioactivity is minimized to maintain the 10

CFR 100 limits for offsite dose consequences in the event of an FHA.

The proposed change to allow penetrations to be unisolated does not

significantly affect the expected dose consequence because of the

absence of containment pressurization potential during fuel movement

or core alterations. The most significant offsite dose contributor

to the fuel handling event is the containment purge system that

generates a motive force for the radioactive material. This flow

path is excluded from the proposed allowance because of this motive

force potential along with flow paths outside the ABSCE. Without

this motive force, as is the case with other penetrations during

fuel movement or core alterations, the potential for additional

offsite dose consequence is unlikely. As an additional measure, this

allowance applies to flow paths that can be filtered by the ABGTS.

Therefore, the margin of safety provided by the containment building

penetration requirements is not significantly impacted by the

proposed allowance to open penetrations under administrative

controls. With the timely provision to identify and isolate affected

penetrations and the provision for ABGTS operability, the margin of

safety is maintained without a significant reduction.

[[Page 1929]]

The NRC 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.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Richard P. Correia.

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

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

Date of application for amendments: November 24, 1999 (TS 99-16).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79

(Unit 2) by updating the Technical Specification (TS) surveillance

requirements for penetration efficiency tests of charcoal adsorbers to

comply with American Society for Testing and Materials (ASTM) test

standard ASTM D3803-1989 as directed by NRC Generic Letter (GL) 99-02.

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 will require laboratory tests of safety-

related charcoal filter adsorbers to tighter specifications. NRC

research indicates that the new test protocols yield more accurate

measures of filter efficiency and better reproducibility of test

results. No physical change is made to the filter by these expanded

timeframes of testing and tighter controls; therefore, no change to

the filter behavior is expected. Current methods for selecting and

obtaining charcoal samples for testing will be retained without

change. The proposed revision does not alter any plant equipment or

operating practices other than filter tests that are conducted away

from the plant site, and as such the probability of an accident is

not increased.

Laboratory test acceptance criteria contain a safety factor to

ensure that the efficiency assumed in the accident analysis is still

valid at the end of the operating cycle. Because ASTM D3803-1989 is

a more accurate and demanding test than older tests, upgrading TSs

to the ASTM D3803-1989 protocol allows use of a safety factor of 2

for determining the acceptance criteria for charcoal filter

efficiency. This safety factor can be used for systems with or

without humidity control because the lack of humidity control is

already accounted for in the test conditions.

Applying the ASTM D3803-1989 test methodology and using the new

safety factor is expected to yield a net improvement in safety. The

ASTM D3803-1989 test protocol is expected to improve the

identification of degraded charcoal filters and lead to their timely

replacement without any adverse effects on filter performance.

Therefore, the change in testing does not significantly increase the

consequences of an accident.

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 in laboratory tests performed on charcoal

filters will not alter plant functions or equipment operating

practices other than possibly resulting in more frequent replacement

of charcoal filters. As stated previously, current methods for

selecting and obtaining charcoal samples for testing will be

retained without change. The ASTM D3803-1989 test methodology is not

expected to alter the filters; therefore, it will not adversely

alter the resulting filter performance. Since the plant functions

and equipment are not altered, the proposed revision will not create

a new or different kind of accident.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

Analyses of design-basis accidents assume a particular ESF

[Engineered Safety Feature] charcoal filter adsorption efficiency

when calculating offsite and control room operator doses. Charcoal

filter samples are tested to determine whether the filter adsorber

efficiency is greater than that assumed in the design-basis accident

analysis. The laboratory test acceptance criteria contains a safety

factor to ensure that the efficiency assumed in the accident

analysis is still valid at the end of the operating cycle. Because

ASTM D3803-1989 is a more accurate and demanding test than older

tests, NRC indicated in GL 99-02 that licensees upgrading their TS

to this new protocol will be able to use a safety factor as low as 2

for determining the acceptance criteria for charcoal filter

efficiency. This safety factor can be used for systems with or

without humidity control because the lack of humidity control is

already accounted for in the test conditions. As stated in the GL,

the new test protocol and associated safety factors have been

reviewed and found to not significantly decrease the margin of

safety.

The NRC 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.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Richard P. Correia.

Previously Published Notices of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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

No. 1, Pope County, Arkansas

Date of amendment request: December 16, 1999.

Brief description of amendment: The proposed change would amend

Technical Specification 4.18.5.b to allow tube 110/60 to remain in

service through the current operating cycle (cycle 16) with two axial

indications that have potential through-wall depths greater than the

plugging limit. The axial indications are located in the roll

transition region and are contained within the upper tubesheet.

Date of publication of individual notice in Federal Register:

December 29, 1999 (64 FR 73080).

Expiration date of individual notice: Comments on no significant

hazards considerations by January 12, 2000; requests for hearing by

January 28, 2000. Clarification: The December 29, 1999, notice

indicated that requests for a hearing with respect to issuance of this

amendment must be filed by January 12, 2000. The correct deadline for

this action is January 28, 2000.

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

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

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, 1999.

Description of amendment request: The proposed changes would revise

the River Bend Station (RBS) Technical Requirements Manual, Section TR

[[Page 1930]]

3.9.14, and add an exception to the current prohibition for travel of

loads in excess of 1200 pounds over fuel assemblies in the spent fuel

storage pool. The exception would allow the licensee to move the spent

fuel pool (SFP) watertight gates, which separate the SFP from the cask

and lower transfer pools, to perform maintenance and repairs on the

gates and watertight seals. Related sections of the RBS Updated Safety

Analysis Report would also be revised to be consistent with the

exception. The licensee determined that movement of the gate, with its

associated rigging, over spent fuel would involve an unreviewed safety

question in accordance with Title 10 of the Code of Federal

Regulations, Section 50.59.

Date of publication of individual notice in Federal Register:

December 21, 1999 (64 FR 71511).

Expiration date of individual notice: January 20, 2000. Correction:

The December 21, 1999, notice indicated that requests for a hearing

with respect to issuance of this amendment must be filed by January 28,

2000. The correct deadline for this action is January 20, 2000.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

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

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 electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.

1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: December 16, 1998, as

supplemented July 16, September 29, and December 21, 1999.

Brief description of amendments: The amendments revise Technical

Specifications 3.8.1 and 3.37 to ensure that the appropriate actions

are taken to prevent double sequencing of safety-related loads and that

the setpoint allowable values for the degraded voltage relays reflect

the required function of the relays.

Date of issuance: December 29, 1999.

Effective Date: December 29, 1999, to be implemented within 90

days.

Amendment Nos: Unit 1-123, Unit 2-123, Unit 3-123.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

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

14279) The July 16, September 29, and December 21, 1999, letters

provided additional clarifying information that was written within the

scope of the original application and Federal Register notice and did

not change the staff's initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated December 29, 1999.

No significant hazards consideration comments received: No.

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

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

Maryland

Date of application for amendments: September 1, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications as follows:

1. Technical Specification 1.1 is changed to replace the definition

of Azimuthal Power Tilt with a new definition.

2. Technical Specification 2.1.1.2 is changed by replacing the peak

linear heat rate safety limit with less than or equal to 22 kW/ft.

3. Technical Specification Surveillance Requirement (SR) 3.3.6.2 is

changed by replacing the degraded voltage function with transient

degraded voltage and steady-state degraded voltage functions.

4. Technical Specification SRs 3.8.1.9 and 3.8.1.15 are changed by

replacing the steady-state voltage range with the range of greater than

or equal to 4060 volts and less than or equal to 4400 volts.

5. Technical Specification 5.6.5.a is changed by adding Technical

Specifications 3.1.4 and 3.3.1 to the list.

6. Technical Specification Figure 2.1.1-1 is changed by removing

the reference to Figure B2.1-1.

7. Various Technical Specifications and Figures 2.1.1-1a are

changed by removing references to Unit 2, Cycle 12, and deleting Figure

2.1.1-1a.

8. Technical Specification 5.6.5.b, Item 41.ii is changed by

correcting CEN-99(B)-P to CEN-119(B)-P.

Date of issuance: December 15, 1999.

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

30 days.

Amendment Nos.: 232 and 208.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 6, 1999 (64 FR

54372).

The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated December 15, 1999.

No significant hazards consideration comments received: No.

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

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

Date of application for amendments: August 6, 1999, as supplemented

on November 15, 1999.

Brief description of amendments: The amendments revised Technical

[[Page 1931]]

Specification 3/4.4.6, ``Vacuum Relief'' to remove specific operability

requirements related to position indication for the suppression

chamber-drywell vacuum breakers. The amendments also reformat the

action statement for inoperable vacuum breakers, increase the

surveillance interval for verifying that the vacuum breakers are

closed, and delete the requirement to verify that the manual isolation

valves are closed for an inoperable and open vacuum breaker.

Date of issuance: December 21, 1999.

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

Amendment Nos.: 138 and 122.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 25, 1999 (64 FR

46428).

The November 15, 1999, submittal provided additional clarifying

information that did not change the staff's initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No.

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

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

Date of application for amendments: July 16, 1999.

Brief description of amendments: The amendments revise Technical

Specification 4.7.D.6 by replacing the leakage limit of 11.5 standard

cubic feet per hour (scfh) for each main steam isolation valve (MSIV)

with a limit of 46 scfh on the total combined leakage for the MSIVs of

all four main steam lines.

Date of issuance: December 21, 1999.

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

Amendment Nos.: 192 and 188.

Facility Operating License Nos. DPR-29 and DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 25, 1999 (64 FR

46429).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No.

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 application for amendments: October 15, 1999.

Brief description of amendments: The amendments revise Section

5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the

Technical Specifications by adding a new paragraph. The existing single

paragraph of Section 5.5.7 requires that inspection of each reactor

coolant pump flywheel be done per the recommendations of Regulatory

Position C.4.b of Regulatory Guide 1.14. The amendments add a new

paragraph which specifies that in lieu of Regulatory Positions C.4.b(1)

and C.4.b(2), alternative inspection techniques may be used. Date of

issuance: December 21, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days from the date of issuance.

Amendment Nos.: 182 (Unit 1); 174 (Unit 2).

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 17, 1999 (64

FR 62705).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No.

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

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

Date of application for amendments: October 15, 1999.

Brief description of amendments: The amendments revise Section

5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the

Technical Specifications by adding a new paragraph. The existing single

paragraph of Section 5.5.7 requires that inspection of each reactor

coolant pump flywheel be done per the recommendations of Regulatory

Position C.4.b of Regulatory Guide 1.14. The amendments add a new

paragraph which specifies that in lieu of Regulatory Positions C.4.b(1)

and C.4.b(2), alternative inspection techniques may be used.

Date of issuance: December 21, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days from the date of issuance.

Amendment Nos.: 190 (Unit 1); 171 (Unit 2).

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 17, 1999 (64

FR 62706).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No

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 18, 1998, as

supplemented September 13, 1999.

Brief description of amendment: This amendment revises the St.

Lucie, Unit 2 (SL-2), Technical Specifications (TS) Index Page III, TS

1.10, Dose Equivalent iodine-131; TS 2.1.1.2, Linear Heat Rate; TS

3.1.1.1/4.1.1.1.1, Shutdown Margin--Tavg Greater than 200

deg.F; TS 3/4.1.1.2, Shutdown Margin--Tavg Less Than or

Equal to 200 deg.F; TS 3.1.2.2, Boration Systems Flow Paths--Operating;

TS 3.1.2.4, Charging Pumps--Operating; TS 3.1.2.6, Boric Acid Makeup

Pumps--Operating; TS 3.1.2.8, Borated Water Sources--Operating; and TS

6.9.1.11, Core Operating Limits Report (COLR). The amendment also

relocates the core operating limits for shutdown margin to the SL-2

COLR. The following Bases have also been changed in connection with

this amendment: TS Bases 2.1.1, Reactor Core; Bases Figure B2.1-1,

Axial Power Distributions for Thermal Margin Safety Limits; TS Bases

2.2.1, Reactor Trip Setpoints (Variable Power Level-High); TS Bases 3/

4.1.1.1 and 3/4.1.1.2, Shutdown Margin; and TS Bases 3/4.1.2, Boration

Systems.

Date of Issuance: December 21, 1999.

Effective Date: As of date of issuance, to be implemented prior to

fuel reload for Cycle 12.

Amendment No.: 105.

Facility Operating License No. NPF-16: Amendment revised the TS.

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6697). The supplemental letter dated September 13, 1999, provided

additional information that did not expand the scope of the amendment

request as noticed or change the original proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No.

[[Page 1932]]

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: April 26, 1999.

Brief description of amendments: The amendments revise the

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

the Technical Specification Index and to remove inconsistencies, and

make administrative changes. A portion of the request, related to the

proposed deletion of dates for the approved security plans, was denied.

Date of issuance: December 20, 1999.

Effective date: December 20, 1999.

Amendment Nos.: 203 and 197.

Facility Operating License Nos. DPR-31 and DPR-41: Amendments

revised the TS.

Date of initial notice in Federal Register: June 2, 1999 (64 FR

29711).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 20, 1999.

No significant hazards consideration comments received: No.

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

Station, Unit 1 Dauphin County, Pennsylvania

Date of application for amendment: December 3, 1998, as

supplemented January 11, February 4, March 4, March 10, and March 15,

1999.

Brief description of amendment: This amendment conforms the license

to reflect the transfer of Facility Operating License No. DPR-50 for

the Three Mile Island Nuclear Station, Unit 1, from GPU Nuclear, Inc.,

et al., to AmerGen Energy Company, LLC, as previously approved by Order

dated April 12, 1999.

Date of issuance: December 20, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 218.

Facility Operating License No. DPR-50: Amendment revised the

license and the Technical Specifications.

Date of initial notice in Federal Register: December 21, 1998 (63

FR 70436).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 12, 1999.

Comments received: Yes. See safety evaluation dated April 12, 1999.

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

County, Iowa

Date of application for amendment: April 12, 1999, as supplemented

October 5 and 8, 1999.

Brief description of amendment: The amendment revises Technical

Specification (TS) Surveillance Requirement (SR) 3.6.1.3.7 to allow a

``representative sample'' of reactor instrumentation line excess flow

check valves (EFCVs) to be tested every 24 months, instead of testing

each EFCV every 24 months.

Date of issuance: December 29, 1999.

Effective date: December 29, 1999.

Amendment No.: 230.

Facility Operating License No. DPR-49: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38028).

The October 5 and 8, 1999, letters provided clarifying information

that was within the scope of the original Federal Register notice and

did not change the staff's initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1999.

No significant hazards consideration comments received: No.

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

Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: December 3, 1998.

Brief description of amendment: This amendment revised the

Technical Specifications for sealed source leakage testing to

specifically address testing requirements for fission detectors.

Date of issuance: December 20, 1999.

Effective date: December 20, 1999, with full implementation within

45 days.

Amendment No.: 235.

Facility Operating License No. DPR-58: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43773).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 20, 1999.

No significant hazards consideration comments received: No.

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: September 17, 1999, as

supplemented November 10 and 19, 1999.

Brief description of amendments: The amendments would approve the

licensee's revision of the Updated Final Safety Analysis Report and

Emergency Operating Procedures to use methodology to credit the

negative reactivity provided by insertion of the rod cluster control

assemblies (RCCAs) into the reactor core following any design basis

loss-of-coolant accident, during realignment from a cold leg

recirculation to a hot leg recirculation configuration. This change to

the licensing basis, when evaluated by the licensee in accordance with

10 CFR 59.59, resulted in an unreviewed safety question that requires

prior approval by the NRC staff in accordance with the provisions of 10

CFR 50.90 prior to implementation. The amendments also change the Bases

for Technical Specifications Section 3/4.5.5, Refueling Water Storage

Tank.

Date of issuance: December 28, 1999.

Effective date: As of the date of issuance and shall be implemented

within 60 days.

Amendment Nos.: 236 and 218.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 20, 1999 (64 FR

56531).

The licensee's letters of November 10 and 19, 1999, provided

additional information that did not change scope of the application or

the staff's proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 28, 1999.

No significant hazards consideration comments received: No.

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: November 5, 1999.

Brief description of amendment: The amendments would revise Unit 1

and 2 Technical Specification (TS) 3.5.1, Action ``a'' and ``b,'' to

reflect the monitoring of pressure from the Reactor Coolant System

instead of the pressurizer. The amendment would also revise Unit 1 and

2 TS Surveillance Requirement 4.5.1.c to require verification that

power is removed from each emergency core cooling system accumulator

isolation valve operator instead of verification that each accumulator

isolation valve breaker is physically removed from the circuit.

Furthermore, the amendment would make administrative changes to Unit 1

and 2 TS Bases 3/4.5.1.

Date of issuance: December 23, 1999.

[[Page 1933]]

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 237 and 219.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 23, 1999 (64

FR 65735).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 23, 1999.

No significant hazards consideration comments received: No.

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

Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: August 17, 1999.

Brief description of amendment: The amendment removes the steam

generator voltage-based repair criteria, F* repair criteria, and

sleeving methodologies from the Unit 1 Technical Specifications and

clarifies the Bases sections accordingly.

Date of issuance: December 22, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days.

Amendment No.: 238.

Facility Operating License No. DPR-58: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: October 6, 1999 (64 FR

54375).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1999.

No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: November 8, 1999.

Brief description of amendment: The amendment changed action

statements, definitions, and footnotes pertaining to the Technical

Specifications for primary containment leakage and primary containment

purge system to allow an alternative approach for isolating a bypass

leakage path and/or purge system line.

Date of issuance: December 16, 1999.

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

30 days.

Amendment No.: 87.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: November 16, 1999 (64

FR 62228).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1999.

No significant hazards consideration comments received: No

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: March 31, 1999.

Brief description of amendment: Amendment changes Technical

Specification Table 3.6.1.2-1 by adding two relief valves, and

associated leak rate criteria, to be installed on the drywell equipment

drain line and drywell floor drain line.

Date of issuance: December 16, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 88.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24197).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1999.

No significant hazards consideration comments received: No

North Atlantic Energy Service Corporation, et al., Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: March 5, 1998.

Description of amendment request: This amendment revises the

Technical Specifications (TSs) by relocating the procedural details of

the Radiological Effluent Technical Specifications (RETS) to the

Offsite Dose Calculation Manual. The TSs were also revised to relocate

procedural details associated with solid radioactive wastes to the

Process Control Program. In addition, the Administrative Controls

section of the TSs was revised to incorporate programmatic controls for

radioactive effluents and environmental monitoring. These 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.''

Date of issuance: December 15, 1999.

Effective date: As of its date of issuance, and shall be

implemented within 120 days.

Amendment No.: 66.

Facility Operating License No. NPF-86: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19972). The Commission received comments which were addressed in the

staff's Safety Evaluation dated December 15, 1999.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 15, 1999.

No significant hazards consideration comments received: Yes.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna

Nuclear Power Plant, Wayne County, New York

Date of application for amendment: October 20, 1999.

Brief description of amendment: The amendment changes from December

31, 1999, to June 30, 2001, the date specified in TS 4.3.1.1.b Note

associated with maintaining spent fuel pool boron concentration >2300

ppm at all times until a permanent resolution to the current

criticality concerns is implemented.

Date of issuance: December 21, 1999.

Effective date: December 21, 1999.

Amendment No.: 75.

Facility Operating License No. DPR-18: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 19, 1999 (64

FR 63345).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: September 4, 1998, as

supplemented on February 8, April 16, August 26, September 16, and

November 17, 1999.

Brief description of amendment: The amendment increases the spent

fuel pool storage capacity from 2,870 to 3,353 fuel assemblies.

Date of Issuance: December 21, 1999.

Effective date: As of the date of issuance, and shall be

implemented within 30 days.

[[Page 1934]]

Amendment No.: 182

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 1, 1998 (64 FR

52774). The supplemental information did not affect the staff's

proposed no significant hazards consideration determination, and was

within the scope of the original amendment application as published.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated December 21, 1999.

No significant hazards consideration comments received: No.

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

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

Virginia

Date of application for amendments: November 18, 1998, as

supplemented by letter dated October 22, 1999.

Brief description of amendments: The amendments change the North

Anna Power Station Technical Specifications (TS) to increase the

allowable groundwater elevation at the southeast section of the service

water reservoir dike from 277 to 280 feet at the toe and from 280 to

295 feet at the crest. In addition, TS Table 3.7-6 has been reorganized

to clarify zones of interest in the Service Water Reservoir, the

location of piezometer devices, and piezometer device numbers. The

proposal to eliminate device numbers from the TS was denied because the

device number helps to indicate the location of the piezometer within

the zone as well as the piezometer itself. Finally the column heading

for Allowable Drain Flow Rate was clarified to be the total flow rate.

Date of issuance: As of the date of issuance and shall be

implemented within 30 days.

Effective date: December 29, 1999.

Amendment Nos.: 220 and 201.

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

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

FR 69349). The supplemental letter dated October 22, 1999, contained

clarifying information only, and did not change the initial no

significant hazards determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 29, 1999.

No significant hazards consideration comments received: No.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: April 12, 1999.

Brief description of amendments: These amendments update references

in the Technical Specifications to information in the updated Final

Safety Analysis Report (FSAR). The update is necessary to reflect

relocation of the referenced information in the updated FSAR.

Date of issuance: December 23, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: Unit 1-192; Unit 2-197.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24204).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 23, 1999.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland this 5th day of January 2000.

For the Nuclear Regulatory Commission.

Suzanne Black,

Acting Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 00-611 Filed 1-11-00; 8:45 am]

BILLING CODE 7590-01-P