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.

[[Page 11960]]

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from February 12, 1999, through February 26,

1999. The last biweekly notice was published on February 24, 1999 (FR

64 PR 9183).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

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

filing of requests for a hearing and petitions for leave to intervene

is discussed below.

By April 9, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

[[Page 11961]]

the Secretary of the Commission, U.S. Nuclear Regulatory Commission,

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

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

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

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

General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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 amendment request: January 29, 1999.

Description of amendment request: The amendments would allow credit

for containment overpressure to assist in providing net positive

suction head (NPSH) for the emergency core cooling system pumps for a

period of greater than 8 hours. The current licensing basis recognizes

credit given only to 8 hours after a design-basis loss-of-coolant

accident and the licensee has determined this to be 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:

Does the change involve a significant increase in the

probability or consequences of an accident as previously evaluated?

The proposed amendment involves the available containment

overpressure (COP) following a design basis loss of coolant accident

(DBA-LOCA) and the resulting NPSH available to the RHR [residual

heat removal] and CS [core spray] pumps. While this change affects

the ability of these pumps to perform their required functions

following a DBA-LOCA, it does not affect the reactor recirculation

piping or the reactor coolant pressure boundary, which are the

initiators of the DBA-LOCA. Therefore, the proposed amendment does

not involve a significant increase in the probability of an accident

previously evaluated.

The consequences of a previously analyzed event are dependent on

the initial conditions assumed for the analysis, the availability

and successful functioning of the equipment assumed to operate in

response to the analyzed event, and the set points at which these

actions are initiated. The proposed change permits limited COP to be

credited in the calculation of available NPSH for the RHR and CS

pumps following a DBA-LOCA.

The proposed change is supported by calculations, which

demonstrates that adequate COP will be available to ensure the RHR

and CS systems will be capable of performing their required safety

functions. Therefore, the proposed amendment does not involve a

significant increase in the consequences of an accident previously

evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed amendment permits limited COP to be credited in the

calculation of available NPSH for the RHR and CS pumps following a

DBA-LOCA. This amendment does not involve a physical alteration of

the plant. The proposed amendment is supported by calculations,

which demonstrate that adequate COP will be available to ensure the

RHR and CS systems will be capable of performing their required

safety functions. This amendment will not alter the manner in which

the RHR and CS systems are initiated, nor will the function demands

on the RHR or CS system be changed. Therefore, the proposed

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

of accident from any previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed amendment permits limited COP to be credited in the

calculation of available NPSH for the RHR and CS pumps following a

DBA-LOCA. Crediting an incremental amount of overpressure does not

result in a significant reduction in the margin of safety, because

conservative analyses demonstrate that adequate COP will be

available to ensure the RHR and CS systems will be capable of

performing their required safety functions. Therefore, the proposed

amendment does not involve a significant reduction in a margin of

safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Project Director: Stuart A. Richards.

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,

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

Date of amendment request: February 18, 1999.

Description of amendment request: The proposed amendments would

revise the joint Technical Specifications (TSs): (1) Surveillance

Requirement (SR) 3.6.16.1--This SR incorrectly characterizes the access

openings (there are five of them) to the reactor building as each

having a double-door design, when in reality there is a single door for

each opening; the proposed revision would change the wording to

correctly characterize the actual design. (2) SR 3.6.16.3--This SR

specifies that the reactor building structural integrity inspection be

performed every 40 months to 50 months and during shutdown; the

proposed revision would change this frequency to three times every 10

years coinciding with containment visual examinations required by SR

3.6.1.1. (3) Administrative Control 5.5.2--The proposed revision would

add wording to specify that containment visual examinations required by

Regulatory Guide c.3 will be conducted three times every 10 years

including during each shutdown for SR 3.6.1.1.

The proposed amendments would only revise the SRs and

Administrative Controls specified above; no physical change to any

plant design is involved.

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:

First Standard

Implementation of this amendment would not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Approval of this amendment will have no

significant effect on accident probabilities or consequences. The

containment and reactor building are not accident initiating systems

or structures; therefore, there will be no impact on any accident

probabilities by the approval of this amendment. The containment and

reactor buildings serve an important function to mitigate

consequences of postulated accidents previously evaluated and the

examination frequencies proposed in this amendment will not result

in a reduction in

[[Page 11962]]

their capacity to meet their intended function. Therefore, there

will be no impact on the consequences of any accident previously

evaluated.

Second Standard

Implementation of this amendment would not create the

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

previously evaluated. No new accident causal mechanisms are created

as a result of NRC approval of this amendment request. No changes

are being made to the plant that will introduce any new accident

causal mechanisms. This amendment request does not impact any plant

systems that are accident initiators, since the containment and

reactor building function primarily as accident mitigators.

Third Standard

Implementation of this amendment would not involve a significant

reduction in a margin of safety. Margin of safety is related to the

confidence in the ability of the fission product barriers to perform

their design functions during and following an accident situation,

including the performance of the containment and reactor building.

These components are already capable of performing as designed, and

their functions are verified by visual examination and leakage rate

testing. The ability of the containment and reactor building to

perform their design function will not be impaired by the

implementation of this amendment at Catawba Nuclear Station.

Consequently, no safety margin will be impacted.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina

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

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

North Carolina.

NRC Project Director: Herbert N. Berkow.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: January 27, 1999.

Description of amendment request: The proposed amendment would

provide a one-time extension of the inspection interval for the Once

Through Steam Generator (OTSG) tubes specified in the Crystal River

Unit 3 (CR-3) Improved Technical Specifications (ITS) to coincide with

the planned operating cycle. CR-3 ITS 5.6.2.10 requires the OTSG

inspection interval to be 24 calendar months for Category C-2

inspection results. However, due to a previous extended maintenance

outage, the next OTSG inspection at CR-3, which is planned for the

October 1999 refueling and maintenance outage, will be approximately 26

calendar months since the last inspection. Florida Power Corporation

indicated that the total interval between inspections would correspond

to less than 21.6 months of plant operation at a temperature of

500 deg.F or above (measured at the hot leg side of the OTSG). The

licensee stated that the conclusions reached in the operational

assessments for the OTSGs show leakage and structural integrity are

maintained by substantial margins until the end of the planned

operating cycle.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The last Crystal River Unit 3 (CR-3) Once Through Steam

Generator (OTSG) tube surveillance was completed in August 1997.

Both standard and enhanced eddy current techniques were used to

inspect 100% of the OTSG tubes. Operational assessments performed

for CR-3 provide reasonable assurance that the OTSG performance

criteria meet the leakage and structural requirements in Draft

Regulatory Guide-1074. These performance criteria will be maintained

until the end of the planned operating cycle. These operational

assessments demonstrate that operation is acceptable for an

operating cycle length of up to 21.6 months of operating time at a

temperature of 500 deg.F or above (measured at the hot leg side).

The operational assessments concluded that the projected

cumulative leakage for the limiting OTSG would be less than 1 gallon

per minute (gpm) under the limiting accident conditions at the end

of the planned operating cycle. Thus, the accident analysis

assumptions bound the condition of the OTSGs, and structural and

leakage integrity will be maintained for the proposed operating

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

significant increase in the probability or consequences of any

accident previously evaluated.

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

from previously evaluated accidents?

No new failure modes or accident scenarios are created by

changing the inspection from a frequency based on calendar months,

to a one-time interval based on up to 21.6 months of operating time

at a temperature of 500'F or above (measured at the hot leg side).

Plant systems and components will not be operated in a different

manner as a result of this change. Thus, this change does not

increase the risk of a plant trip or present a challenge to any

other safety system. For all known degradation mechanisms in the CR-

3 OTSGs, the most recent operational assessments bound the

probability of tube burst and project primary-to-secondary leakage

at accident conditions for the end of Operating Cycle 11 to be less

than 1 gpm. Therefore, the proposed one-time change does not create

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

previously evaluated.

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

Improved Technical Specification (ITS) Bases 3.4.12 contains

relevant information pertaining to the limitations on reactor

coolant system (RCS) leakage. The ITS Bases discuss the 1 gpm

primary-to-secondary leakage assumed for a main steam line break

accident, as well as for a steam generator tube rupture accident.

The evaluation provided by this license amendment request shows that

tube structural integrity is maintained, thus the required

structural margins specified in NRC Regulatory Guide 1.121 are

satisfied. The operational assessments performed show the maximum

accident leakage, assuming all these indications leak, is less than

1 gpm. Therefore, all known OTSG tube degradation mechanisms have

been assessed, and the proposed one-time change does not involve a

significant reduction in the margin of safety.

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

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

satisfied.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Project Director: Cecil O. Thomas.

Northeast Nuclear Energy Company (NNECo), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: January 18, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Table 3.7-6, ``Area Temperature

Monitoring,'' by increasing the temperature limits for the fuel

building fuel pool pump cubicles and fuel building general area. The

amendment would also change the Millstone Unit 3 licensing basis by

incorporating into the Millstone Unit 3 Final Safety Analysis Report

(FSAR) a revision to describe the full core off-load condition as a

normal evolution. In

[[Page 11963]]

addition, the amendment would increase the maximum bulk spent fuel pool

(SFP) temperature from 140 deg. F to 150 deg. F, allow the crediting of

evaporative cooling as a decay heat removal mechanism for the SFP (use

of the ONEPOOL computer code), and allow the use of Holtec's quality

assurance validated DECOR computer code as a method for predicting

decay heat loads in the SFP pool.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10 CFR 50.92, NNECo has reviewed the proposed

changes and has concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10 CFR 50.92(c) are not

[satisfied]. The proposed changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed license amendment will permit NNECo to conduct full

core off-loads as a normal evolution through the end of plant life.

This amendment request does not affect: (1) the number of spent fuel

assemblies allowed in the spent fuel pool, (2) Spent Fuel Pool (SFP)

criticality analysis, (3) structural analysis of the spent fuel pool

or (4) radiological release scenarios.

The proposed license amendment permits the use of ORIGEN2 based

DECOR and ONEPOOL codes for the analysis of the Unit 3 SFP. The

ORIGEN2 based DECOR code more accurately predicts decay heat loads

from the spent fuel in the SFP. The ONEPOOL code credits the effect

of evaporative cooling on the SFP bulk temperature. The use of these

codes improves the accuracy of predicting SFP bulk temperatures

during normal and abnormal refueling scenarios.

The analysis of decay heat removal permits the discharge of fuel

from the reactor vessel to the SFP [to] start as early as 132 hours

(depending on cooling water temperature) after reactor shutdown at a

rate of 3 assemblies per hour. The existing accident analysis for a

dropped spent fuel bundle during refueling bounds this situation as

the analysis assumed a decay time of 100 hours after reactor

shutdown.

The increase in pool temperature from 140 deg. F to 150 deg. F

does not significantly impact the structural integrity of the fuel

handling equipment. The temperature increase does not create a new

failure of the fuel handling equipment that has not been previously

analyzed.

The increased SFP temperature results in higher ambient

temperatures in the Fuel Building. However, the duration of an

increased pool temperature event is limited. The effect on the

environmental qualification (EQ) of electrical equipment is an

increase in the Maximum Normal and Abnormal Excursion temperatures,

which are based on short duration excursions from the predicted

summer maximum temperatures. This is reflected in the proposed

Technical Specification (TS) temperature changes. The temperature

limits within TS, 3.7.14, ``Plant Systems: Area Temperature

Monitoring,'' Table 3.7-6, for the Fuel Pool Pump Cubicles and Fuel

Pool General Area increase from 110 deg. F to 119 deg. F, and from

104 deg. F to 108 deg. F respectively, based upon the revised

environmental conditions. The proposed TS changes do not involve a

significant increase in the probability or consequences of an

accident previously analyzed as the Fuel Building Ventilation System

is qualified for the increased temperature and humidity conditions.

There are no changes in the EQ of equipment.

A comprehensive review of the design of the SFP, Spent Fuel Pool

Cooling and Purification system and other associated systems,

structures and components has been completed. All systems,

structures and components are fully qualified at the higher SFP

temperature of 150 deg. F for a full core off-load as a normal

operation.

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

significant increase in the probability or consequence of an

accident previously analyzed.

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

from any previously evaluated.

The proposed license amendment will permit NNECo to conduct full

core off-loads as a normal evolution through the end of plant life.

There are no physical plant changes. The SSCs [systems, structures,

and components] supporting the SFP and Spent Fuel Pool Cooling are

fully qualified for operation at 150 deg. F. The higher Fuel Pool

Pump Cubicles and Fuel Pool General Area temperatures do not create

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

previously evaluated. Thus the changes do not create the possibility

of an accident of a different type than previously evaluated.

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

The proposed license amendment will permit NNECo to conduct full

core off-loads as a normal evolution through the end of plant life.

The proposed changes allow a higher heat load in the SFP which

results in a higher calculated maximum temperatures than the current

analysis. In addition, several changes have been made with respect

to the analysis methods used in calculating the maximum

temperatures.

The new analysis demonstrates that the SFP cooling configuration

will maintain the SFP pool bulk temperature at or below 150 deg. F

with a single train of spent fuel pool cooling. This temperature is

above the SRP [Standard Review Plan] guidance of 140 deg. F but is

well below the 212 deg. F limit permitted for abnormal core off-

loads as defined in the Standard Review Plan (NUREG -0800). This

temperature guideline of 140 deg. F was one of the acceptance

criteria credited by the NRC staff during their review of the

adequacy of the design of the SFP Cooling System within the NRC

Safety Evaluation Report (SER) for Millstone Unit 3 (NUREG-1031) and

consequently requires prior review and approval.

A single active failure will cause the loss of one of the two

trains of spent fuel pool cooling. The complete loss of cooling to

the Spent Fuel Pool is not a creditable occurrence in that the Fuel

Pool Cooling System is designed to be able to withstand the worst

single failure and still be able to perform its intended function.

However, a loss of cooling analysis indicates that several hours are

available during a refueling, and over thirteen hours are available

during normal operations for operators to respond to the loss of

cooling prior to the Spent Fuel Pool reaching its structural design

temperature of 200 deg. F.

A comprehensive review of the design of the SFP, Spent Fuel Pool

Cooling and Purification System and other associated systems,

structures and components has been completed for qualification at

the higher pool temperature of 150 deg. F. All systems, structures

and components are fully qualified at the higher Technical

Specification Fuel Pool Pump Cubicles and Fuel Pool General Area

temperatures, and at the increased SFP temperature, and are

therefore qualified for a full core off-load as a normal operation.

The ORIGEN2 based DECOR code more accurately predicts decay heat

loads from the spent fuel in the SFP. The ONEPOOL code credits the

effect of evaporative cooling on the SFP bulk temperature. The use

of these codes improves the accuracy of predicting SFP bulk

temperatures during normal and abnormal refueling scenarios. The use

of these computer codes as a method for predicting decay heat loads

and crediting evaporative cooling as a decay heat removal mechanism

have not previously been evaluated for Unit 3, and therefore,

require[s] prior NRC review and approval.

Therefore, based on the above, this license amendment to permit

NNECo to conduct full core off-loads as a normal evolution, increase

the maximum SFP pool bulk temperature from 140 deg. F to 150 deg. F,

use the ORIGEN2 based DECOR and ONEPOOL computer codes to calculate

the decay heat load and determine the effects of evaporative cooling

respectively, and increase the TS Fuel Pool Pump Cubicles and

General Area temperatures, does not involve a significant reduction

in the margin of safety.

Thus, it is concluded that the proposed amendment does not

involve a significant reduction in the margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed amendment does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

[[Page 11964]]

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

Connecticut.

NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: February 10, 1999.

Description of amendment request: The proposed amendment would

incorporate alternative inspection requirements into Technical

Specification Surveillance Requirement 3/4.4.10, ``Structural

Integrity,'' for the reactor coolant pump flywheel.

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:

NNECO has reviewed the proposed revision in accordance with 10

CFR 50.92 and has concluded that the revision does not involve a

Significant Hazards Consideration (SHC). The basis for this

conclusion is that the three criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does not involve a SHC because the

revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

This proposed revision to the Millstone Unit No. 3 Technical

Specifications incorporates alternative reactor coolant pump

flywheel inspection requirements into Surveillance 4.4.10 based on

Topical Report WCAP-14535A. WCAP-14535A provided a technical basis

for the elimination of inspection requirements for reactor coolant

pump flywheels based on industry data. The industry data indicated

that no indications that would affect the integrity of flywheels was

[sic] revealed during 729 examinations of 217 flywheels at 57 plants

(including Millstone Unit No. 3). The NRC, during their review and

approval of the WCAP required continued inspections on a ten year

interval to protect against events and degradation that were not

anticipated and had not been considered in the WCAP analysis. The

proposed alternate inspection requirements are consistent with the

conclusions of an NRC review and generic approval of Topical Report

WCAP-14535A. Thus, it is concluded that the proposed revision does

not significantly increase the probability of an accident.

Additionally, the performance of reactor coolant pump flywheel

surveillances does not increase the consequence of an accident

previously evaluated.

Therefore, it is concluded that the proposed revision does not

involve a significant increase in the probability or consequence of

an accident previously evaluated.

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

from any accident previously evaluated.

This proposed revision to the surveillance does not change the

operation of any plant system or component during normal or accident

conditions. The proposed change incorporates alternate inspection

requirements for the reactor coolant pump flywheels that were

generically approved for use by licensees by the NRC. This change

does not include any physical changes to the plant.

Thus, this proposed revision does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

This proposed revision to the Millstone Unit No. 3 Technical

Specifications incorporates alternative reactor coolant pump

flywheel inspection requirements into Surveillance 4.4.10 that are

consistent with the conclusions of an NRC review and generic

approval of Topical Report WCAP-14535A. The current inspection

requirements of Surveillance 4.4.10 and the NRC review of WCAP-

14535A were both based on the recommendations of Regulatory Guide

1.14.

Thus, it is concluded that the proposed revision does not

involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

Connecticut.

NRC Project Director: Elinor G. Adensam.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: February 5, 1999.

Description of amendment requests: The proposed amendments would

modify the technical specifications (TS) to incorporate Revision 3 of

the ABB Combustion Engineering, Inc.'s topical report, CEN-629-P,

``Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using

Leaktight Sleeves'', dated September 1998 (proprietary and

nonproprietary documents available). The current TS requires that steam

generator tube repair using the Combustion Engineering Inc.'s welded

sleeves shall be in accordance with the methods and criteria described

in Revision 2 of CEN-629-P and Addendum 1, Revision 1 of CEN-629-P.

Incorporation of Revision 3 of CEN-629-P would involve the following TS

changes: (1) editorial/administrative change to TS.4.12.D.3 to reflect

adoption of Revision 3 of CEN-629-P, and deletion of reference to

Addendum 1, Revision 1 of CEN-629-P since Revision 3 incorporates

Addendum 1, Revision 1 of CEN-629-P; (2) changes in sleeve installation

practices that incorporate improvements gained by prior experiences;

and (3) more restrictive change to the sleeve repair limit as specified

in TS.4.12.D.1.(f) from 31 percent of the nominal sleeve wall thickness

to 25 percent.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

evaluated.

Editorial changes have no effect on probability or consequences

of accidents previously evaluated. Changes in installation practices

incorporate improvements gained by experience in installing sleeves.

Further, the changes in the installation practices will change

neither the final configuration of installed sleeves nor the post-

installation NDE [nondestructive examination] from that which is

already approved. Accident induced steam generator tube leakage is

not [a]ffected by these changes. Post installation non-destructive

examination will be conducted using VT, UT, and ET as previously

licensed. The changes in repair limits have [led] to repair limits

that are more conservative than those which have been previously

approved. Thus, none of these changes will create the possibility of

a new or different kind of accident from any accident previously

analyzed.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident previously analyzed.

Editorial changes cannot create the possibility of a new or

different kind of accident. Changes in installation practices

incorporate improvements gained by experience in installing sleeves.

Further, changes in installation practices do not change the final

configuration of installed sleeves from that which is already

approved. The changes in repair limits have [led] to repair limits

that are more conservative than those which have been previously

approved.

[[Page 11965]]

Thus, none of these changes will create the possibility of a new or

different kind of accident from any accident previously analyzed.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

Editorial changes have no effect on the margin of safety.

Changes in installation practices incorporate improvements gained by

experience in installing sleeves. Further, changes in installation

practices do not change the final configuration of installed sleeves

from that which is already approved. The changes in repair limits

have [led] to repair limits that are more conservative than those

which have been previously approved. None of these changes will

affect the tube plugging assumptions used in the PINGP [Prairie

Island Nuclear Generating Plant] accident analyses. Thus, none of

these changes will reduce the margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: Cynthia A. Carpenter.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: January 15, 1999.

Description of amendment request: The proposed changes revise

calibration requirements for the local power range monitors (LPRM).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed amendment would not involve a significant hazards

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

1. involve a significant increase in the probability or

consequences of an accident previously evaluated.

This change proposes to remove the listed requirement for the

method of calibration of the LPRM Signal from TS Table 4.1-2 because

the definition for Instrument Channel Calibration provides the

necessary guidance.

Other changes to the bases and adopting signal calibration

frequency units of MWD/T [Megawatt Days per Ton] vice effective full

power hours is consistent with STS [Standard Technical

Specification].

The proposed changes do not increase the probability of an

accident because the proposed surveillance requirements still ensure

that the LPRM signal is adequately calibrated. The proposed change

provides assurance that the associated Reactor Protection System

(RPS) functions are tested consistent with the analysis assumptions.

As a result, the consequences of an accident are not affected by

this change. This change will not alter assumptions relative to the

mitigation of an accident or transient event. Therefore, this change

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes will not physically alter the plant. As

such, no new or different types of equipment will be installed. The

methods governing normal plant operation and testing are consistent

with current safety analysis assumptions. Therefore, this change

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

accident from any accident previously evaluated.

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

The proposed change removes specific calibration method

information in Table 4.1-2 regarding the LPRM signal which is

adequately addressed in the definition for Instrument Channel

Calibration.

Other changes to the Bases and adopting a signal calibration

Frequency units of MWD/T vice effective full power hours is

consistent with STS.

The proposed changes still provide the necessary control of

testing to ensure operability of the RPS instrumentation. The safety

analysis assumptions will still be maintained, thus no question of

safety exists. Therefore, this change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of amendment request: February 2, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 5.6, ``Fuel Storage, Criticality,''

to change the maximum unirradiated fuel assembly enrichment value for

new fuel storage from 4.5 to 5.0 weight percent Uranium-235 and to

allow the use of equivalent criticality control to that provided by the

current TS requirement of 2.35 mg of Boron-10 per linear inch loading

in the Integral Fuel Burnable Absorber (IFBA) pins.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

(a) Fuel Assembly Drop.

There is no increase in the probability of a fuel assembly drop

accident because the mass of a fuel assembly does not increase when

the fuel enrichment is increased. This amendment affects only the

isotopic composition within the fuel pellets of a fuel assembly

without involving any changes to the outward physical

characteristics or structural integrity of the assembly.

The radiological consequences of a new fuel assembly drop

accident do not increase as a consequence of the proposed change to

new fuel enrichment. Because it has not been irradiated, there are

no significant radiological consequences associated with fresh fuel.

The radiological consequences of an irradiated fuel assembly drop

were previously evaluated and approved in the Spent Fuel license

amendment numbers 151/131 (Units 1 & 2 respectively).

(b) Misplaced Fuel Assembly in New Fuel Storage Vault or Spent

Fuel Storage Racks.

There is no increase in the probability of a misplaced fuel

assembly in the New Fuel Storage Vault or Spent Fuel Storage Racks.

The proposed change does not alter the physical structure of the New

Fuel Storage Vault or the Spent Fuel Storage Racks. All new fuel

assembly movements will continue to be made in accordance with

approved procedures.

There is no increase in the consequences of misplacing a fuel

assembly in the new fuel storage racks. The normally-dry new fuel

vault Keff is very small (approximately 0.65), as such,

there is sufficient reactivity margin to the 0.95 limit to bound any

possible misplacement. The double contingency principle does not

require consideration of a second unlikely event. Since a misplaced

bundle constitutes the first unlikely event, presence of moderator

in the normally dry

[[Page 11966]]

new fuel storage racks (a second unlikely event) is not assumed in

evaluating the event.

The inadvertent misplacement of a fresh fuel assembly in the

spent fuel storage racks has the potential for exceeding the

limiting reactivity, should there be a concurrent and independent

accident condition resulting in the loss of all soluble boron.

Administrative procedures to assure the presence of soluble boron

during fuel handling operations will preclude the possibility of the

simultaneous occurrence of the two independent accident conditions.

The analyses supporting Amendments 151/131 demonstrated that 600 ppm

of soluble boron is adequate to compensate for a mis-loaded fuel

event, while plant procedures require the concentration to be

maintained at least 2300 ppm. The proposed change to allow reduced

IFBA B-10 loading does not invalidate these prior analyses since

equivalent reactivity hold down to the 2.35 mg/linear inch B-10

loading will be maintained.

(c) Introduction of Moderator to the New Fuel Vault

There is no increase in the probability of any accident

involving moderator introduction to the new fuel storage vault. The

proposed change affects only the enrichment within the fuel

assemblies. No other plant systems or components are affected by

this change.

There is no increase in the consequences of introducing a

moderator to the new fuel storage vault resulting from increased

fuel enrichment. The new fuel storage vault has been analyzed for

storage of fuel assemblies with nominal enrichments of 4.65 w/o

U235 at the fully flooded condition and 5.00 w/o

U235 at the optimum moderation condition, as described in

the attached Criticality Analysis (Attachment 2). As long as the

requirement for the number of IFBA pins versus assembly enrichment

is met, calculated Keff (including uncertainties and

biases) does not exceed 0.95 under full density conditions and does

not exceed 0.98 under optimum moderation conditions.

These analyses demonstrate that 5.0 w/o enrichment fuel storage

in the New Fuel Storage Vault complies with criticality acceptance

criteria for all moderation conditions. Therefore, based on the

conclusions of the above analyses, the proposed changes will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

accident from any accident previously analyzed.

The proposed Technical specification changes do not involve any

physical changes to the plant or any changes to the method in which

the plant is operated. No physical changes to the new fuel or spent

fuel storage racks are required, nor any changes in the process or

procedures to place fuel in the racks. The enrichment limits and

reactivity hold-down requirements ensure that the assumptions used

in the criticality analyses remain bounding. As such, these changes

do not affect the performance or qualification of safety-related

equipment. Therefore, the possibility of a new or different type of

accident than previously considered i[s] not created.

3. Do not involve a significant reduction in a margin of safety.

The new fuel storage vault has been analyzed for storage of fuel

assemblies with nominal enrichments of 4.65 w/o U235 at

the fully flooded condition and 5.00 w/o U235 at the

optimum moderation condition, as described in the attached

Criticality Analysis (Attachment 2). As long as the requirement for

the number of IFBA pins versus assembly enrichment in Equation 1 is

met, calculated Keff (including uncertainties and biases)

does not exceed 0.95 under full density conditions and does not

exceed 0.98 under optimum moderation conditions.

For the 5.00 w/o U235 enrichment requested, Equation

2, which bounds Equation 1, will be used in the Technical

Specifications related to new fuel storage.

Therefore, since the calculated values of Keff have

been shown to be below the regulatory limits (including

uncertainties and biases) and because they reflect a substantial

subcritical configuration under adverse conditions, the proposed

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

margin of safety.

Previous analyses provided in support of Amendments 151/131

demonstrate that the addition of new fuel having IFBA pins with a

loading of 2.35 mg B-10 per linear inch to the spent fuel racks does

not result in a reduction in the margin of safety. Thus, providing

for reactivity hold down for IFBA pins which is equivalent to a

nominal 2.35 mg B-10/linear inch loading in fresh fuel in the spent

fuel storage racks maintains the current margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

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

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

NRC Project Director: Elinor G. Adensam.

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

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

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

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

Date of amendment request: February 5, 1999.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TSs) to incorporate some of the

generic changes to the Improved Technical Specifications that have been

previously approved by the NRC. In addition, a TS has been added that

would test the Unit 1 automatic scram relay on a periodic basis.

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 the consequences of a previously evaluated event

for the following reasons:

Proposed Change One

The majority of primary containment isolation valves (PCIVs)

should be in the closed position following an accident to prevent

the release of radiation to the environment. Locked PCIVs are

verified to be in the closed position prior to being locked.

Therefore, it is unnecessary for these valves to be verified closed

under the provisions of Surveillance Requirements (SRs) 3.6.1.3.2

and 3.6.1.3.3. The fact that the valves are secured closed assures

they will be in the safe position following an accident.

Furthermore, per Plant Hatch procedure, locked valves are

periodically verified to be in their correct position. This provides

additional assurance the valves will remain in the correct position.

For these reasons, the proposed change does not involve a

significant increase in the probability or the consequences of a

previously evaluated event.

Proposed Change Two

This proposed change does not affect the function of the control

rods, the control rod drive (CRD) system, or the control rod

housing. Thus, the probability of the control rod drop accident

(CRDA) is not increased. Also, this change does not affect the

function of the rod worth minimizer (RWM). As with the present

Technical Specification, no control rods will be moved (via SRs

3.1.3.2 and 3.1.3.3) when below the low power setpoint (LPSP) to

limit interference with respect to the RWM's function in limiting

the consequences of a CRDA. Additionally, no other systems designed

to prevent or mitigate the consequences of any other transient or

accident are affected.

Proposed Change Three

This proposed change merely deletes a redundant specification in

the control rod operability section. The requirement to electrically

disarm an inoperable withdrawn control rod ensures the validity of

banked position withdrawal sequence (BPWS) is maintained, thus

ensuring the mitigation of the consequences of the CRDA. This

proposed change in no way affects the BPWS, the RWM, or the

structures of the control rods and control rod drive. Thus, the

probability, or the consequences, of a previously evaluated event

are not increased by this proposed change.

Proposed Change Four

Any physical deterioration of a station service battery that can

cause degradation of

[[Page 11967]]

battery performance will result in failure of the SR, with the

ensuing inoperable declaration of the battery. A determination that

battery performance is not degraded, or will not degrade, will

result from evaluation of the particular abnormality found while

performing the Surveillance. This is the intent of the Technical

Specification as clarified in the Bases.

Accordingly, the safety function of the station service

batteries is not compromised as a result of this proposed change.

Thus, the consequences of a previously evaluated event are not

affected by this proposed revision. The proposed revision does not

affect any system needed to prevent the occurrence of previously

analyzed events; therefore, the probability of occurrence of a

previously evaluated event is not increased.

Proposed Change Five

The purpose of the primary containment air interlock is to

provide access to the primary containment while maintaining

containment integrity. Extending the Surveillance Frequency on the

airlock to once per 24 months will not increase the likelihood of

occurrence of any previously evaluated event, since no change in the

operation or testing of any system designed for the prevention of

accidents and transients is being made.

Extending the Frequency of the airlock interlock Surveillance

does not increase the consequences of any accident or transient,

since the proposed change does not affect any system designed to

mitigate the consequences of a previously analyzed event. In fact,

the extended Frequency will challenge the airlock interlock less;

thus, the likelihood of a loss of primary containment integrity will

decrease.

Proposed Change Six

This proposed change to the Safety Function Determination

Program (SFDP) description in LCO [Limiting Condition for Operation]

3.0.6 is more restrictive than the existing version. Requiring an

SFDP evaluation upon entry into LCO 3.0.6, as stated in the Bases,

will not increase the probability of occurrence or the consequences

of a previously evaluated event, since this is purely an

administrative change to clarify the intent of LCO 3.0.6 and provide

consistency with the Bases.

Proposed Change Seven

This proposed administrative change merely relocates the review

requirements for the Offsite Dose Calculation Manual (ODCM) to

licensee controlled documents. This change does not affect any

system designed for the prevention or mitigation of previously

analyzed events or any assumptions regarding transient and accident

analyses.

Proposed Change Eight

This proposed administrative change eliminates some of the

redundant reporting requirements for safety limit violations listed

in the Technical Specifications. This change does not affect any

systems designed for the prevention or mitigation of any previously

evaluated accident or transient. Additionally, the change does not

affect any assumptions of previously evaluated accidents or

transient analyses.

Proposed Change Nine

This change adds a footnote to Unit 1 Technical Specifications

Table 3.3.1.1-1 to ensure the auto scram relays (K14s) are tested as

part of the manual scram Functional Test. This change does not

adversely affect the ability of the reactor protection system (RPS)

to perform its safety function. In fact, the added testing

requirement enhances the ability to detect and correct problems with

the RPS. Successful testing of the K14s on a weekly basis for many

years has demonstrated that the additional testing requirements do

not impose an undue burden on the system. No other systems designed

for the prevention or mitigation of accidents are affected by this

change. Therefore, the probability, or the consequences, of a

previously evaluated event are not increased.

2. The proposed changes do not create the possibility of an

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

Proposed Change One

Removing the SR to verify locked valves are in their ``safe''

position does not increase the likelihood of occurrence or

consequences of a new type of event, since no new modes of operation

are introduced. All plant systems will continue to be operated

within their design basis. Since the valves are verified to be in

their safe position prior to locking, and are periodically verified

to be in that position per the locked valve procedure, the valves

will be in the position assumed by accident analyses should an event

occur.

Proposed Change Two

This proposed change does not affect the function of either the

CRD system or the RWM. These systems, as well as all other systems

designed for the prevention or mitigation of accidents, will

continue to function per their design basis. Also, the BPWS will

continue to be used for control rod withdrawal. Thus, no new modes

of operation that would cause a type of failure different from any

previously analyzed are introduced.

Proposed Change Three

Deleting Required Action B.1 of Technical Specification 3.1.3

does not eliminate any Required Actions, since the subject Required

Action is redundant. Deleting the redundant specification does not

prevent any of the control rod control systems from performing their

functions per their design bases. Therefore, no new modes of

operation are introduced, and the probability of a new type event is

also not introduced by this proposed change.

Proposed Change Four

No changes to the operation, maintenance, or testing of the

batteries are proposed. The batteries will continue to operate

within their design basis. As a result, no new modes of operation

are introduced, and thus, the probability of occurrence of a new

type event is not created.

Proposed Change Five

This change is administrative in the sense that it does not

result in the airlock being operated or tested outside of its

design. The proposed revision only includes a change to the

Frequency of SR 3.6.1.2.2, which tests the interlock's ability to

prevent the two primary containment airlock doors from opening at

the same time. This change does not affect how the test is to be

performed or how the doors are operated. Therefore, the probability

of occurrence of a new type event is not increased by the proposed

change.

Proposed Change Six

This proposed administrative change to the SFDP description does

not involve the operation of any safety-related system. Furthermore,

this change does not involve accident or transient analyses; thus,

no changes to the assumptions for the analyses are made. As a

result, the probability of occurrence of a new type event is not

increased.

Proposed Change Seven

This administrative change merely relocates the review

requirements for the ODCM to licensee controlled documents. This

change does not affect any system designed for the prevention or

mitigation of previously analyzed events or any assumptions

regarding transient and accident analysis. Accordingly, the

possibility of a new type event is not created.

Proposed Change Eight

This administrative change eliminates some of the redundant

reporting requirements for safety limit violations listed in the

Technical Specifications. This change does not affect any systems

designed for the prevention or mitigation of any previously

evaluated accident or transient. Additionally, the change does not

affect any assumptions of previously evaluated accident or transient

analyses. Accordingly, the possibility of a new type event is not

created.

Proposed Change Nine

Adding a requirement to test the auto scram relays (K14s) on a

weekly basis does not create a new mode of operation for the RPS.

Also, no other safety-related systems are affected by this change,

and as a result, the possibility of occurrence of a new type

accident is not created.

3. The changes do not significantly reduce the margin of safety.

Proposed Change One

Not requiring position surveillance on PCIVs locked in position

does not reduce the margin of safety, because the valves are

verified to be in their ``safe'' position prior to locking. This

ensures the valve will remain in the ``safe'' position until it is

unlocked again. The position of these locked valves is verified

periodically by the Operations Department. Furthermore, a

``malicious'' unlocking of the valves is unlikely to take place,

since the keys to the valves are controlled by the shift supervisor

(SS). Anyone wanting to check out a key must obtain SS approval.

Also, the locked valves are periodically verified to be in their

proper

[[Page 11968]]

position whenever Operations Management deems it necessary. For

these reasons, the margin of safety is not significantly reduced.

Proposed Change Two

Moving the Technical Specification 3.1.3 Note from the Required

Action column to the Completion Time column will not affect the

safety function of the RWM system. The RWM will continue to function

through the power ranges where the control rod drop accident is of

concern. The change does not affect the safety function of the RWM

in any way. Thus, the margin of safety is not reduced.

Proposed Change Three

This proposed change only eliminates a redundant Specification.

Adherence to the requirements of the BPWS will still be maintained

during plant startups. Also, the operation of the RWM system remains

unaffected by this proposed change. For these reasons, the margin of

safety for the CRDA is not reduced.

Proposed Change Four

This proposed change clarifies that the purpose of SR 3.8.4.3 is

to determine whether a physical deterioration that could affect

battery performance exists. This is already stated in the Plant

Hatch Technical Specifications Bases; thus, the proposed revision is

merely a clarification of the Specification. Adding this

clarification does not reduce the margin of safety with respect to

battery performance, because an engineering evaluation must be

performed to document that the particular deficiency will not

prevent the battery from performing its safety function.

Proposed Change Five

This proposed change to extend the Frequency of SR 3.6.1.2.2

reduces the number of challenges to primary containment integrity.

The nature of the Surveillance is such that the primary containment

(drywell) interlock is challenged. With that challenge, the

likelihood of a primary containment breach is increased. Therefore,

reducing the Frequency of this SR actually increases the safety of

margin, since normal entry and exit procedures do not permit

challenging the interlock.

Proposed Change Six

This purely administrative change clarifies the definition of

the SFDP in LCO 3.0.6. The Technical Specifications margin of safety

is enhanced, since the new wording, together with the existing

wording in the Bases, makes it clear that the SFDP must be performed

any time LCO 3.0.6 is entered.

Proposed Change Seven

This proposed change merely allows relocation of the review and

approval functions for the ODCM revisions from the Technical

Specifications to owner-controlled documents. The purely

administrative change does not affect any Technical Specifications

required system, test, or function. Changes to the ODCM will

continue to receive the level of review necessary to ensure any

proposed changes are accurate and complete. Therefore, the margin of

safety is not reduced.

Proposed Change Eight

This purely administrative change eliminates redundant reporting

requirements with respect to a safety limit violation. The change

has no effect on any Technical Specifications required system, test,

or function, or on any other safety-related system. Accordingly, the

margin of safety is not reduced.

Proposed Change Nine

This proposed change ensures the Unit 1 auto scram relays (K14s)

are tested on a weekly basis. General Electric recognizes this as an

optimum test frequency for these scram contactors. In this respect,

the margin of safety is increased, since this change ensures the

relays will be tested at the optimum recommended Frequency. Also, at

Plant Hatch, the K14 relays and contacts have been tested at this

Frequency for many years. As a result, placing this requirement on

the relays will not pose an undue burden on the RPS. No other

safety-related systems are affected by this proposed change. For the

above reasons, this proposed change does not reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

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

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

NRC Project Director: Herbert N. Berkow.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas.

Date of amendment request: January 26, 1999.

Description of amendment request: The amendment would revise part

of the Inservice Inspection requirements for the Reactor Coolant Pump

flywheel from an in-place ultrasonic volumetric examination of the

areas of higher stress concentration at the bore and keyway at

approximately 3-year intervals and a surface examination of all exposed

surfaces and complete ultrasonic volumetric examination at

approximately 10 year intervals to ultrasonic examination over the

volume from the inner bore of the flywheel to the circle of one-half

the outer radius once every 10 years.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change increases the examination volume and revises

the periodicity of the ultrasonic examination required by Regulatory

Guide 1.14 regulatory position C.4.b(1) from 3-year intervals to 10-

year intervals. This change is consistent with the conclusions of

the NRC staff in the referenced safety evaluation of WCAP-14535. The

NRC staff has determined that the evaluation methodology is

appropriate and the criteria are in accordance with the design

criteria of RG 1.14. There is no change in the method of plant

operation or system design.

The proposed change revises the inspection process to eliminate

10-year surface examination of all exposed surfaces and complete

ultrasonic volumetric examination required by Regulatory Guide 1.14

Regulatory Position C.4.b(2). An ultrasonic volumetric examination

will be performed of a section of the flywheel once every 10 years.

This change is consistent with the conclusions of the NRC staff in

referenced safety evaluation of WCAP-14535. The NRC staff has

determined that the evaluation methodology is appropriate and the

criteria are in accordance with the design criteria of RG 1.14.

Based on the above, this change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change increases the examination volume and revises

the periodicity of the ultrasonic examination required by Regulatory

Guide 1.14 regulatory position C.4.b(1) from 3-year intervals to 10-

year intervals. This change is consistent with the conclusions of

the NRC staff in the referenced safety evaluation of WCAP-14535. The

only potential accident associated with this change is loss of the

flywheel. Precautionary measures taken to preclude missile formation

from Reactor Coolant Pump components assure that the pumps will not

produce missiles under any anticipated accident condition. Each

component of the primary pump motors has been analyzed for missile

generation Any fragments of the motor rotor would be contained by

the heavy stator. Effects on reactor coolant flow due to loss of

functionality of a single Reactor Coolant Pump flywheel are

enveloped by the analysis of the consequences of the Reactor Coolant

Pump locked rotor event. There is no change in the method of plant

operation or system design.

The proposed change revises the inspection process to eliminate

10-year surface examination of all exposed surfaces and complete

ultrasonic volumetric examination required by Regulatory Guide

[[Page 11969]]

1.14 Regulatory Position C.4.b(2). An ultrasonic volumetric

examination will be performed of a section of the flywheel once

every 10 years. This change is consistent with the conclusions of

the NRC staff in the referenced safety evaluation of WCAP-14535. The

only potential accident associated with this change is loss of the

flywheel. Precautionary measures taken to preclude missile formation

from Reactor Coolant Pump components assure that the pumps will not

produce missiles under any anticipated accident condition. Each

component of the primary pump motors has been analyzed for missile

generation. Any fragments of the motor rotor would be contained by

the heavy stator. Effects on reactor coolant flow due to loss of

functionality of single Reactor Coolant Pump flywheel are enveloped

by the analysis of the consequences of the Reactor Coolant Pump

locked rotor event.

Based on the above, this change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change increases the examination volume and revises

the periodicity of the ultrasonic examination required by Regulatory

Guide 1.14 Regulatory Position C.4.b(1) from 3-year intervals to 10-

year intervals. This change is consistent with the conclusions of

the NRC staff in the referenced safety evaluation of WCAP-14535. The

NRC staff used deterministic methodology to review the WCAP and came

to the conclusion that ASME margins would be maintained during the

service period and a 10-year inspection period appears reasonable.

There is no change in the method of plant operation or system

design.

The proposed change revises the inspection process to eliminate

the 10-year surface examination of all exposed surfaces and complete

ultrasonic volumetric examination required by Regulatory Position

C.4.b(2) of Regulatory Guide 1.14. An ultrasonic volumetric

examination will be performed of a section of the flywheel once

every 10 years. This change is consistent with the conclusions of

the NRC staff in the referenced safety evaluation of WCAP-14535.

Effects on reactor coolant flow due to loss of functionality of a

single Reactor Coolant Pump flywheel are enveloped by the analysis

of the consequences of the Reactor Coolant Pump locked rotor event.

Based on the above, 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: John N. Hannon.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,

Unit 2, Hamilton County, Tennessee

Date of application for amendments: August 27, 1998 (TS 98-04).

Brief description of amendments: The proposed amendment would

change the Sequoyah (SQN) Technical Specifications (TSs) by adding a

provision to Section 5.3, ``Reactor Core,'' authorizing a limited

number of lead test assemblies (LTAs) to be installed in the core as

described in the Framatome Cogema Fuels Report BAW-2328 entitled

``Blended Uranium Lead Test Assembly Design Report.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), Tennessee Valley

Authority (TVA), 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 LTAs are identical to the other Mark-BW fuel assemblies with

the exception of minor differences internal to the fuel rods. These

differences will not adversely affect reactor neutronic or thermal-

hydraulic performance; therefore, they do not significantly increase

the probability of accidents while in the reactor.

The reload design analyses performed for SQN Unit 2 Cycle 10

accounts for any minor neutronic differences of the LTAs and

confirms any effects on the reload core to be within established

fuel design limits.

The pressure and temperature safety limits for the cycles in

which the LTAs will be in the core are the same as those for the

current operating cycle thus ensuring that the fuel will be

maintained within the same range of safety parameters that form the

basis for the FSAR [Final Safety Analysis Report] accident

evaluation. The potential effects of the LTAs on plant operation and

safety have been evaluated. This evaluation investigated both LOCA

[loss-of-coolant accident] and non-LOCA events, and concluded that

the current analyses remain bounding and that there will be no

increase in the probability of occurrence for any design basis

accident described in the FSAR.

The impact of the LTAs on key safety analysis parameters was

examined and it was concluded that there will be an insignificant

impact.

The impacts of the LTAs on the radiological consequences for all

postulated events have been evaluated. The total calculated source

term and the source term-activity of isotopes, which significantly

contribute to operator and off-site accident exposure levels, were

shown to be less than standard fuel assemblies, therefore, it will

not increase the consequences of any accident previously evaluated.

The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The fuel assembly design for the LTAs is identical to the

standard fuel assemblies. The main difference between the LTAs and

the production fuel is that the concentration of the U234

and U236 isotopes will be higher in the LTA fuel pellets

than that typically found in standard fuel. These isotopic

differences will not affect the chemical, mechanical or thermal

properties of the fuel pellet.

The LTAs meet the same design criteria and licensing basis

criteria as the standard fuel assemblies and were manufactured with

the same processes. The LTA skeleton is identical to the standard

skeleton, which ensures that the loadings associated with normal

operation, seismic events, LOCA events, and shipping and handling

are not affected.

Pressure and temperature safety limits will be maintained the

same as those for the current operating cycle, thus ensuring that

the fuel will be maintained within the same range of safety

parameters that form the basis for previous accident evaluations. No

new performance requirements are being imposed on any system or

component that exceed design criteria or cause the core to operate

in excess of design basis operating limits. No credible scenario has

been identified, which could jeopardize equipment that could cause

intensify or mitigate events or accident sequences. Therefore, the

LTAs will not create the possibility of accidents or equipment

malfunctions of a different type than previously evaluated while in

the reactor.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The LTAs will not adversely affect reactor neutronic or thermal-

hydraulic performance. The LOCA acceptance criteria with LTAs

installed in the core will continue to be met: peak cladding

temperature of less than or equal to 2200 deg.F, peak cladding

oxidation of less than or equal to 17 percent, average clad

oxidation of less than or equal to 1 percent, and long-term

coolability. The acceptance criteria for departure from nucleate

boiling (DNB) events with the LTAs installed in the core will also

continue to be met: 95 percent probability and 95 percent confidence

interval that DNB is not occurring during the transient. Other

acceptance criteria have also been demonstrated to remain within

acceptable limits. The total calculated source term-activity and the

source term-activity of isotopes, which significantly contribute to

operator and off-site accident exposure levels of the LTAs, was

determined to be less than that for the standard fuel assembly. All

previously evaluated events remain bounding and valid. For these

reasons, the proposed amendment does not involve a significant

reduction in a margin of safety.

[[Page 11970]]

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.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Cecil O. Thomas.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

Date of application for amendments: August 14, 1998, as

supplemented by letters dated October 13, 1998, and December 23, 1998.

Brief description of amendments: The amendments revised the Dresden

Technical Specifications (TS) to reflect the use of Siemens Power

Corporation ATRIUM-9B fuel. Specifically the amendments incorporated

the following into the TS: (a) new methodologies that enhanced

operational flexibility and reduced the likelihood of future plant

derates; (b) administrative changes that eliminated the cycle-specific

implementation of ATRIUM-9B fuel and adopted Improved Standard

Technical Specification language where appropriate; and (c) changed the

Minimum Critical Power Ratio.

Date of issuance: February 16, 1999

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

Amendment Nos.: 171; 166.

Facility Operating License Nos. DPR-19 and DPR-25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: 63 FR 48258 (September

9, 1998) and 63 FR 59588 (November 4, 1998). The October 13 and

December 23, 1998 submittals provided additional clarifying information

that did not change the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

Date of application for amendments: October 16, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) to lower the power level (from 30 percent

to 25 percent rated thermal power) below which the turbine control

valve (TCV) and turbine stop valve (TSV) closure scram signals and the

end-of-cycle recirculation pump trip (EOC-RPT) signal are not in

effect. The amendments also (1) delete from TSs the reference to

turbine first stage pressure as a measure of rated thermal power, and

(2) add a requirement to periodically verify that TCV and TSV scram

trip functions and the EOC-RPT trip functions are not bypassed at

greater than or equal to 25 percent rated thermal power.

Date of issuance: February 12, 1999.

Effective date: For Unit 1--Immediately, to be implemented within

90 days; for Unit 2--immediately, to be implemented prior to startup of

L2C8.

Amendment Nos.: 130; 114.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 54108). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

Commonwealth Edison Company, Docket No. 50-374, LaSalle County Station,

Unit 2, LaSalle County, Illinois

Date of amendment request: November 9, 1998.

Brief description of amendment: The amendment revised Technical

Specification 3/4.3.2, ``Isolation Actuation Instrumentation'' to add/

revise various isolation setpoints for leak detection instrumentation.

These changes are necessary due to modifications to the reactor water

cleanup (RWCU) system to restore ``hot'' suction to the RWCU pumps and

due to a re-evaluation of the high energy line break analysis. In

addition, the amendment eliminated isolation actuation trip functions

for the residual heat removal system steam condensing mode and shutdown

cooling mode.

Date of issuance: February 16, 1999.

Effective date: February 16, 1999.

Amendment No.: 115.

Facility Operating License No. NPF-18: The amendment revised the

Technical Specifications.

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

FR 69335).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois

[[Page 11971]]

Valley Community College, Oglesby, Illinois 61348-9692.

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

Michigan

Date of application for amendment: June 20, 1997 (NRC-97-0037), as

supplemented July 2, 1997 (NRC-97-0066), and March 10 (NRC-98-0036) and

April 9, 1998 (NRC-98-0083).

Brief description of amendment: The amendment revises the technical

specifications by relocating surveillance requirement 4.4.1.1.2 for

setting the reactor recirculation system motor-generator set scoop tube

stops to the updated final safety analysis report (UFSAR), with

modifications.

Date of issuance: February 8, 1999.

Effective date: February 8, 1999, with full implementation within

90 days. Implementation of this amendment shall include the relocation

of surveillance requirement 4.4.1.1.2 from the technical specifications

to the UFSAR as described in the licensee's application dated June 20,

1997, as supplemented on July 2, 1997, and March 10 and April 9, 1998,

and evaluated in the staff's safety evaluation dated February 8, 1999.

Amendment No.: 130.

Facility Operating License No. NPF-43: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38134)

The July 2, 1997, and March 10 and April 9, 1998, submittals

provided additional clarifying information 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 February 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: June 30, 1998, as supplemented

by letter dated November 23, 1998.

Brief description of amendment: The amendment authorizes the

licensee to modify the plant to correct a design deficiency with the

plant protection system (PPS). This deficiency could have rendered the

system vulnerable to a single failure (i.e., failure of a DC buss) with

one channel in bypass. The proposed modification would ensure the

required redundancy and independence for the PPS such that no single

failure results in a loss of the protection function with a channel in

indefinite bypass, and removal from service of any component or channel

does not result in a loss of the minimum redundancy required by the

Technical Specifications.

Date of issuance: February 17, 1999.

Effective date: This license amendment is effective as of its date

of issuance to be implemented within six months following the

facility's restart from refueling outage 2R14.

Amendment No.: 201

Facility Operating License No. NPF-6: Amendment revised the license

to authorize a modification to the plant protection system.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66593).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 29, 1998, as supplemented by letter

dated January 12, 1999.

Brief description of amendment: The amendment changes the Appendix

A TSs by modifying TS 3.7.6.1, ``Control Room Emergency Air Filtration

System'' in Modes 1-4, TS 3.7.6.2, ``Control Room Emergency Air

Filtration System'' in Modes 5 and 6, TS 3.7.6.3, ``Control Room Air

Temperature'' in Modes 1-4, TS 3.7.6.4, ``Control Room Air

Temperature,'' in Modes 5 and 6, TS 3.7.6.5, ``Control Room Isolation

and Pressurization,'' and its associated basis. This amendment also

modifies TS Tables 3.3-6 and 4.3-3 for the Control Room Intake

Monitors.

Date of issuance: February 17, 1999.

Effective date: This license amendment is effective as of its date

of issuance, to be implemented within 60 days.

Amendment No.: 149.

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

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

56247).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: May 28, 1996, as supplemented by

letter dated October 27, 1998.

Brief description of amendment: This amendment increases the test

interval for reactor protection system instrumentation and anticipatory

reactor trip system instrumentation.

Date of issuance: February 22, 1999.

Effective date: February 22, 1999.

Amendment No.: 230.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 31, 1996 (61 FR

40031). The supplemental information provided did not impact the

proposed no significant hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Library, Government Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606

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

Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: September 8, 1997, as

supplemented by submittal dated October 27, 1998.

Brief description of amendment: This amendment revised Technical

Specification 5.2.2.e, ``Organization--Unit Staff,'' by removing the

reference to the NRC Policy Statement on working hours and

incorporating a requirement for administrative procedures necessary to

ensure that the working hours of unit staff who perform safety-related

functions are limited and controlled.

Date of issuance: February 22, 1999.

Effective date: February 22, 1999.

Amendment No.: 98.

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 19, 1997 (62

FR

[[Page 11972]]

61847) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

Power Plant, Unit 1, Lake County, Ohio

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

supplemented by submittals dated December 3, and December 9, 1998 and

January 12, and January 26, 1999.

Brief description of amendment: This amendment revised Technical

Specification 3.8.1, ``AC Sources--Operating,'' by extending the

emergency diesel generator (EDG) Completion Time from 72 hours to 14

days for the Division 1 and 2 EDG and allows performance of the EDG 24-

hour test run in Modes 1 and 2. The amendment also establishes

Technical Specification 5.5.13.1, ``Configuration Risk Management

Program,'' an administrative program that assesses risk based on plant

status.

Date of issuance: February 24, 1999.

Effective date: February 24, 1999.

Amendment No.: 99.

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

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

56261)

The supplemental information provided clarifying information that

did not change the initial no significant hazards consideration

determination or alter the scope of the proposed action.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

County, Iowa

Date of application for amendment: January 22, 1999.

Brief description of amendment: The amendment revises Technical

Specification Surveillance Requirement (SR) 3.8.1.7 to better match

plant conditions during diesel generator (DG) testing by clarifying

which voltage and frequency limits are applicable during the transient

and steady state portions of the DG start. A Notice of Enforcement

Discretion (NOED) related to SR 3.8.1.7 was issued verbally on January

20, 1999. The NOED is documented in a letter dated January 22, 1999.

Date of issuance: February 17, 1999.

Effective date: February 17, 1999, to be implemented within 30

days.

Amendment No.: 225.

Facility Operating License No. DPR-49: Amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration (NSHC): Yes (64 FR 4902 dated

February 1, 1999). The notice provided an opportunity to submit

comments on the Commission's proposed NSHC determination. No comments

have been received. The notice also provided for an opportunity to

request a hearing by March 3, 1999, but indicated that if the

Commission makes a final NSHC determination, any such hearing would

take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of

exigent circumstances, and final NSHC determination are contained in a

Safety Evaluation dated February 17, 1999.

Attorney for Licensee: Al Gutterman; Morgan, Lewis & Bockius, 1800

M Street NW, Washington, D.C. 20036-5869.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: October 22, 1998.

Brief description of amendment: The amendment revises Technical

Specifications 3.3.2.1, ``Instrumentation--Engineered Safety Feature

Actuation System Instrumentation''; 3.4.9.3, ``Reactor Coolant System--

Overpressure Protection Systems''; and 3.5.3, ``Emergency Core Cooling

Systems--ECCS Subsystems--Tavg < 300 [degrees] F.'' The amendment

allows Millstone Unit No. 2 to prevent an automatic start of any high-

pressure safety injection (HPSI) pump when the shutdown cooling system

(SDCS) is in operation (Mode 4 and below). An inadvertent start of an

HPSI pump could result in overpressurization of the SDCS.

Date of issuance: February 10, 1999.

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

60 days from the date of issuance.

Amendment No.: 227.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66600)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 10, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

California

Date of application for amendments: March 10, 1997, as supplemented

by letters dated May 20, 1997; March 13, August 28, and October 22,

1998; and January 29 and February 2, 1999.

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant (DCPP) Unit Nos. 1 and 2 that changed TS 3/4.4.5 and its

associated Bases to allow the implementation of steam generator (SG)

tube alternate repair criteria for axial indications in the

Westinghouse explosive tube expansion (WEXTEX) region below the top of

the tubesheet and below the bottom of the WEXTEX transition that may

exceed the current TS depth-based plugging limit.

Date of issuance: February 19, 1999.

Effective date: February 19, 1999, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 1--129; Unit 2--127.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 19, 1997 (62

FR 61843). The March 13, August 28, and October 22, 1998; and January

29 and February 2, 1999, supplemental letters provided additional

clarifying information, did not expand the scope of the application as

originally noticed, and did not change the staff's original proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated February 19, 1999. No

[[Page 11973]]

significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

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

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

Date of application for amendment: June 16, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) by moving certain administrative requirements from

the TSs to the Final Safety Analysis Report.

Date of issuance: February 25, 1999.

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

30 days.

Amendment No.: 188.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 29, 1998 (63 FR

40560).

No significant hazards consideration comments received: No.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit

2, Luzerne County, Pennsylvania

Date of application for amendment: August 4, 1998, as supplemented

by letters dated December 16, 1998, and January 12 and 28, 1999.

Brief description of amendment: This amendment would modify the

Susquehanna Steam Electric Station, Unit 2 Technical Specifications to

replace figures 2.1.1.2-1 and 2.1.1.2-2, and associated footnotes, with

single value minimum critical power ratio Safety Limits of Section

2.1.1.2; remove references from Section 5.6.5 which do not directly

support the generation of Core Operating Limits; remove references from

Section 5.6.5 which were previously included to address the application

of the ANFB-10 correlation to ATRIUM-10 fuel; include Siemiens Power

Corporation ANFB-10 topical report in Section 5.6.5; and to change the

Bases to reflect the inclusion of the ANFB-10 critical power

correlation.

Date of issuance: February 17, 1999.

Effective date: As of date of issuance, to be implemented in 30

days.

Amendment No.: 154.

Facility Operating License No. NPF-22. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48262). The December 16, 1998, and January 12, and 28, 1999, letters

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: October 12, 1998.

Brief Description of amendments: The amendments revise Technical

Specification Section 6, ``Administrative Controls,'' to recognize the

additional management positions associated with the steam generator

replacement project. The new positions would provide the ability to

approve procedures regarding this project, which may affect nuclear

safety.

Date of issuance: February 19, 1999.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1--141 and Unit 2--133.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64122). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of application for amendments: June 19, 1998, as supplemented

by letters dated December 4, 1998, and January 13, 1999.

Brief description of amendments: The proposed changes would modify

the technical specifications (TS) to (1) reduce the minimum RCS cold

leg temperature (Tc); (2) convert the specified reactor coolant system

(RCS) flow from mass units (lbm/hr) to volumetric units (gpm); and (3)

eliminate the maximum RCS flow rate limit from the TS.

Date of issuance: February 12, 1999.

Effective date: February 12, 1999, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2--149; Unit 3--141.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48266). The supplemental letters dated December 4, 1998, and January

13, 1999, provided additional clarifying information, did not expand

the scope of the application as originally noticed, and did not change

the staff's original proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of application for amendments: November 23, 1998, as

supplemented by letter dated January 13, 1999.

Brief description of amendments: The amendments revised the

technical specifications (TS) to (1) reinstate the log power reactor

trip at or above 4E-5% RATED THERMAL POWER (RTP); (2) reinstate reactor

trips for Reactor Coolant Flow--Low (RCS flow), the Local Power

Density--High (LPD), and the Departure from Nucleate Boiling Ratio--Low

(DNBR); (3) remove the word ``automatically'' from notes (a) and (d) of

Table 3.3.1-1 to clarify that the

[[Page 11974]]

manual enable of the trip is permissible; and (4) clarify that the

setpoints on Table 3.3.1-1 are set relative to logarithmic power.

Date of issuance: February 12, 1999.

Effective date: February 12, 1999, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2--150; Unit 3--142.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998

(63 FR 71973). The January 13, 1999, supplemental information

provided additional clarifying information and did not change the

staff's initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: November 23, 1998.

Brief description of amendments: Relocates descriptive design

information from Technical Specification (TS) Section 3.7.1.1, Table

3.7-2, regarding orifice sizes for main steam line Code safety valves,

to the Bases section for this TS.

Date of issuance: February 24, 1999.

Effective date: This license amendment is effective as of its date

of issuance, and shall be implemented within 30 days of issuance.

Amendment Nos.: Unit 1--Amendment No. 103; Unit 2--Amendment No.

90.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71974).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: December 10, 1998.

Brief description of amendment: The amendment corrects an error in

the technical specifications by changing to the use of ``hydrogen

balance air'' rather than the incorrect ``hydrogen balance nitrogen''

for calibration of the Augmented Offgass System hydrogen monitors.

Date of Issuance: February 12, 1999.

Effective date: February 12, 1999, to be implemented within 30

days.

Amendment No.: 166.

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

Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71975).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated February 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 20, 1997, as

supplemented on September 18, 1997, and October 31, 1997.

Brief description of amendment: The amendment makes administrative

changes to the Technical Specifications to add and revise reference to

NRC-approved methodologies which will be used to generate the cycle-

specific thermal operating limits in the Vermont Yankee Core Operating

Limits Report.

Date of Issuance: February 23, 1999.

Effective date: February 23, 1999, to be implemented within 30

days.

Amendment No.: 167.

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

Technical Specifications.

Date of initial notice in Federal Register: March 25, 1998 (63 FR

14489).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated February 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: December 10, 1996, as

supplemented on January 22, 1999.

Brief description of amendment: The amendment makes changes to the

Technical Specifications regarding fire protection requirements as

recommended by NRC Generic Letters 86-10 and 88-12. This includes

relocating certain fire protection requirements to the Vermont Yankee

Fire Protection Plan, Technical Requirements Manual, and Final Safety

Analysis Report.

Date of Issuance: February 24, 1999.

Effective date: February 24, 1999, to be implemented within 30

days.

Amendment No.: 168.

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

Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register: February 26, 1997 (62

FR 8801).

The January 22, 1999, supplement did not change the original

proposed no significant hazards consideration determination, or expand

the scope of the amendment request as initially noticed.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated February 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-

281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: November 4, 1998.

Brief Description of amendments: These amendments revise the

Technical Specifications (TS) to change Emergency Diesel Generator

start and load time testing requirements in TS 4.6.A.1.b. The TS Basis

Section 3.16 is also revised to reflect the basis for the new TS

requirements. The TS changes are in a conservative direction, and are

being made to bring the TS and the Updated Final Safety Analysis Report

into conformance with each other.

Date of issuance: March 1, 1999.

Effective date: March 1, 1999.

Amendment Nos.: 218 and 218.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

change the Technical Specifications.

[[Page 11975]]

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4161).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: March 4, 1998, as supplemented

September 21, 1998.

Brief description of amendment: This amendment revises the

Technical Specifications to provide a one-hour limiting condition for

operation that will permit a safety injection pump to be used for the

addition of make-up fluid to safety injection accumulators during power

operation.

Date of issuance: February 23, 1999.

Effective date: February 23, 1999.

Amendment No.: 143.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17237).

The September 21, 1998, supplement provided clarifying information

that did not change the initial no significant hazards determination or

alter the scope of the proposed action.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Dated at Rockville, Maryland, this 3rd day of March 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-5751 Filed 3-9-99; 8:45 am]

BILLING CODE 7590-01-P