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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 21, 1999, through June 4, 1999. The last

biweekly notice was published on June 2, 1999 (64 FR 29707).

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 July 19, 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

[[Page 32285]]

proceeding, but such an amended petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of amendments request: May 23, 1997, as revised by letters

dated September 27, 1998, and May 26, 1999.

Description of amendments request: The proposed amendments would

revise Technical Specification (TS) Limiting Condition of Operation

(LCO) 3.4.14 and TS Sections 5.5.9 and 5.6.8 to allow the use of steam

generator (SG) tube sleeves as an alternative to plugging defective SG

tubes. The May 26, 1999, letter completely revised the May 23, 1997,

request for amendments, and this notice supersedes the original Federal

Register notice dated July 30, 1997 (62 FR 40845).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change to TS LCO 3.4.14.d and e will replace the

leakage limits of 1 gallon per minute (gpm) primary to secondary

leakage through all SGs and 720 gallon per day (gpd) through any one

SG with a new limit of 150 gpd through any one SG. This is a more

restrictive change. A TS limit of 150 gpd primary to secondary

Leakage through any one steam generator is significantly less than

the initial conditions assumed in the safety analyses. The 150 gpd

limit is based on operating experience as an indication of one or

more propagating tube leak mechanisms. The Steam Generator Tube

Surveillance Program described in TS Section 5.5.9 ensures that the

structural integrity of the SG tubes is maintained. The leakage rate

limit of 150 gpd for any one SG provides additional assurance

against tube rupture at normal and faulted conditions and provides

additional assurance that cracks will not propagate to burst prior

to detection by leakage monitoring methods and commencement of plant

shutdown. Therefore, this change to TS LCO 3.4.14.e will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes to TS 5.5.9 will add inservice inspection

requirements for SG tube sleeves. These requirements will ensure

that all installed SG tube sleeves will be inspected prior to

initial operation and routinely thereafter, to assure the capability

of each sleeve to perform its design function during each operating

cycle. The tube sleeves will be the Combustion Engineering, Inc. (CE

or ABB-CE) Leak Tight sleeves, as described in CE report CEN-630-P,

``Repair of \3/4\'' O.D. Steam Generator Tubes Using Leak Tight

Sleeves,'' Revision 02, dated June 1997. (This proprietary report is

provided as Enclosure 4 with this submittal.) The tube sleeve

dimensions, materials and joints are designed to the applicable ASME

[American Society of Mechanical Engineers] Boiler and Pressure

Vessel code requirements. An extensive test program was performed

that demonstrated that the sleeves will fulfill their intended

function as leak tight structural members. Evaluation of sleeved

tubes indicates no detrimental effects on the sleeve-tube assembly

resulting from reactor coolant system flow, coolant chemistries, or

thermal and pressure conditions. Structural analyses of the sleeve-

tube assembly have established its integrity under normal and

accident conditions. Mechanical testing using ASME code stress

allowables was performed to support the analyses. Also, corrosion

tests were performed and revealed no evidence of sleeve or tube

corrosion considered detrimental under anticipated service

conditions. A sleeved tube will exhibit greater hydraulic resistance

and reduced heat transfer capability than an un-sleeved tube.

However, these effects are much less than would be imposed by taking

the tube out of service by plugging. Section 10.0 of CE report CEN-

630-P describes the analyses to determine the hydraulic and heat

transfer effects. Calculations using plant-specific information will

identify sleeve-to-plug equivalency ratios. The proposed changes to

the SG inservice inspection program will assure that sleeved SG

tubes will meet the structural requirements of tubes that are not

defective. The proposed sleeve plugging limit of 35% of nominal wall

will ensure that the sleeves remaining in service will perform their

design function. Also, installation of

[[Page 32286]]

sleeves will not significantly [a]ffect the primary system flow rate

or the heat transfer capability of the SGs. Therefore, this change

to TS section 5.5.9 will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The change to the SG reporting requirements in TS section 5.6.8

will ensure that the number of sleeved SG tubes will be reported to

the NRC along with the number of plugged tubes. This is an

administrative change that has no effect on the operation or

maintenance of the plant and will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

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

different kind of accident from any accident previously evaluated?

The proposed change to TS LCO 3.4.14.d and e will replace the

leakage limits of 1 gpm primary to secondary leakage through all SGs

and 720 gpd through any one SG with a new limit of 150 gpd through

any one SG. This is a more restrictive change that will provide

added assurance against steam generator tube ruptures. Since the

current allowable primary to secondary leakage is being reduced,

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

kind of accident from any accident previously evaluated.

The proposed changes to TS section 5.5.9 for the SG inservice

inspection program will assure that sleeved SG tubes will meet the

structural requirements of tubes that are not defective. Also,

installation of sleeves will not significantly [a]ffect the primary

system flow rate or the heat transfer capability of the SGs.

Therefore, this change to TS section 5.5.9 will not create the

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

previously evaluated.

The change to the SG reporting requirements in TS section 5.6.8

will ensure that the number of sleeved SG tubes will be reported to

the NRC along with the number of plugged tubes. This is an

administrative change that has no effect on the operation or

maintenance of the plant and will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

margin of safety?

The proposed change to TS LCO 3.4.14.d and e will replace the

leakage limits of 1 gpm primary to secondary leakage through all SGs

and 720 gpd through any one SG with a new limit of 150 gpd through

any one SG. This is a more restrictive change that will provide

added assurance against steam generator tube ruptures. Since the

current allowable primary to secondary leakage is being reduced,

this change will not involve a significant reduction in a margin of

safety.

The proposed changes to TS section 5.5.9 for the SG inservice

inspection program will assure that sleeved SG tubes will meet the

structural requirements of tubes that are not defective. Also,

installation of sleeves will not significantly [a]ffect the primary

system flow rate or the heat transfer capability of the SGs.

Therefore, this change to TS section 5.5.9 will not involve a

significant reduction in a margin of safety.

The change to the SG reporting requirements in TS section 5.6.8

will ensure that the number of sleeved SG tubes will be reported to

the NRC along with the number of plugged tubes. This is an

administrative change that has no effect on the operation or

maintenance of the plant and will not involve a significant

reduction in a margin of safety.

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

that review, it appears that the three standards of 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: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Section Chief: Stephen Dembek.

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

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

Date of amendment request: May 5, 1999.

Description of amendment request: The proposed amendments would

revise the basis for evaluation of the reactor building ventilation

(VR) system exhaust plenum masonry walls. Specifically, the amendment

would approve the use of different methodology and acceptance criteria

for the reassessment of certain masonry walls subjected to transient

pressurization loads resulting from a high energy line break.

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 previously evaluated?

The change involves reassessment of the VR exhaust plenum due to

a transient pressurization during a Main Steam Line Break (MSLB).

Since the transient pressurization is a result of the MSLB, and the

block walls and the dampers are not initiators of any accident, the

probability of an accident previously evaluated is not affected.

This analysis does not affect the total amount of radioactive

release due to the MSLB Outside of the Primary Containment, so the

total offsite dose consequences does not change. A small portion of

the release, which passes the dampers prior to closure, will now be

an elevated release via the plant ventilation stack instead of a

ground level release. The original analysis assumed the entire

release was a ground level release, and thus remains bounding for

the MSLB accident.

The Control Room and Auxiliary Electric Equipment Room (AEER)

dose consequences are impacted only slightly due to the small amount

of steam/air mixture released from the new pressure relief damper.

The steam/air mixture becomes mixed with the air volume in that area

of the Auxiliary Building but was all assumed to be available for

inleakage to the Control Room and AEER. The dose increase for the

Control Room and AEER is less than or equal to 0.05 Rem thyroid and

negligible change to the whole body dose, such that the dose due to

the MSLB accident remains much less than the DBA LOCA dose and

General Design Criteria 19. The MSLB accident dose consequences

remain bounded by the Design Basis Loss of Coolant Accident.

The effects of the steam released by the pressure relief damper

into the Auxiliary Building has been evaluated for environmental

qualification impact on systems, structures and components (SSCs) in

the area of the Auxiliary Building affected for both radiation and

steam/temperature affects. The effect on area temperature is about 4

deg.F and is above initial temperature for not more than 24 hours.

The change in humidity is negligible, and radiation dose impact is

small and bounded by previous calculations.

These consequences assume that the VR exhaust plenum masonry

walls do not rupture based on the design changes being made in

conjunction with the masonry wall reevaluation for each LaSalle Unit

that will prevent the failure of the VR exhaust plenum masonry

walls.

Therefore this proposed amendment does not involve a significant

increase in the probability or 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 MSLB accident is previously analyzed but considered only

instantaneous closure of installed dampers. The reevaluation and

design changes extend the previous accident analysis to assure that

structures previously considered unaffected by the MSLB will

maintain their structural integrity. The block walls are static and

the dampers function in response to an accident, thus the analysis

method and design changes are not accident initiators. Therefore the

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

of accident from any accident previously evaluated.

The design changes being made in conjunction with the masonry

wall reevaluation for each LaSalle Unit that will prevent the

failure of the VR exhaust plenum masonry walls are as follows:

(1) Installation of a pressure relief damper,

(2) An excess-flow check damper, and

(3) Required masonry wall support improvements in the reactor

building ventilation exhaust plenum for each Unit.

The reevaluation of the masonry walls uses different load

factors and load combinations

[[Page 32287]]

as well as reduced acceptance criteria than previously used for

these walls. The change in the evaluation does not cause the rupture

or failure of the effected masonry walls, since the evaluation shows

the walls remain intact.

The installation of the above design changes, in conjunction

with masonry wall analysis assure that the subject masonry walls

will not rupture or fail. Therefore, SSCs that would be affected by

wall rupture can fulfill their intended function, maintaining the

consequences of previously evaluated accident the same.

The new pressure relief damper and excess-flow check damper are

safety-related and are analyzed to function under the conditions

created by the MSLB. In addition, the dampers and the duct they are

installed in have been analyzed to assure no failure will occur

during an Operating Basis Earthquake (OBE) or Safe Shutdown

Earthquake (SSE).

Based on an analysis of potential failure modes in accordance

with ANSI/ANS-58.9-1981, ``Single Failure Criteria for Light Water

Reactor Safety-Related Fluid Systems,'' Paragraph 4.1, the active

function of the pressure relief damper and excess flow check damper

are considered exempted from consideration of single failure. The

principles governing operation of the dampers are simple and direct

and not subject to change or deterioration with time, similar to the

function of a code safety relief valve and a swing check valve. With

periodic testing of the dampers, continued reliable performance is

assured.

The dampers are designed and set so that the pressures created

by normal ventilation flow changes do not cycle the dampers, and

thus the new dampers do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Administrative controls will be in place prior to implementation

of this change to assure the testing and maintenance is periodically

performed in accordance with vendor recommendations. These dampers

will be included as equipment required to be monitored/maintained,

because the function performed by the dampers is within the scope of

the Maintenance Rule, 10 CFR 50.65.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

Does the change involve a significant reduction in a margin of

safety?

Originally, no masonry walls were evaluated for HELB

pressurization effects, because the walls were considered protected

by the isolation dampers. However, the original design methodology

for masonry did include load combinations including Pa:

Abnormal

1.0D + 1.0L + 1.5Pa

Abnormal/Severe Environment

1.0D + 1.0L + 1.25Pa + 1.25Eo

Abnormal/Extreme Environment

1.0D + 1.0L + 1.0Pa + 1.0Ess,

Where D is Dead Load; L is Live Load; Pa is

pressurization due to HELB; Eo is Loads generated by the

Operating Basis Earthquake (OBE); and Ess is Loads

generated by the Safe Shutdown Earthquake (SSE).

The current reevaluation was required due to determination that

some block walls in the LaSalle Auxiliary Building are affected by a

transient pressurization due to a MSLB. The specific changes from

the original analyses involve the following for loads and load

combinations.

1. Abnormal:

1.0D + 1.0L + 1.0PHELB

2. Abnormal/severe environmental:

1.0D + 1.0L + [(1.1Eo)2 +

1.0PHELB2]\1/2\

3. Abnormal/extreme environmental:

1.0D + 1.0L + [1.0Ess2 +

1.0PHELB2]\1/2\

Where:

(1) PHELB is the short-term differential pressurization

load on the VR plenum masonry walls resulting from non-instantaneous

opening/closure of the protection dampers.

(2) The Load Factor on pressure due to HELB is 1.0 for all cases.

(3) The Loading Combination of pressure and seismic is the Square

Root of the Sum of Squares (SRSS).

LaSalle has selected the proposed load combinations in

consideration of the following:

Isolation, check, and relief dampers protect the walls;

therefore the pressurization effects are not sustained, but are

transient in nature.

The transient pressurization effect (PHELB) is

derived from a conservative detailed analysis of an instantaneous

HELB combined with non-instantaneous damper opening/closure. Due to

the precise nature and conservatism of this HELB analysis, there is

little uncertainty in PHELB .

Therefore a load factor of 1.0 is used for all abnormal load

combinations.

PHELB is a short duration, dynamic load. Accordingly,

the seismic and transient HELB pressurization loads are combined

using the Square Root of Sum of the Squares (SRSS) method because

the peak effects of these dynamic loads are unlikely to occur

simultaneously. This combination method is used in the analysis of

other components such as component supports.

The proposed load combinations accordingly provide a

conservative basis for reassessment of the VR exhaust plenum masonry

wall systems.

In regards to the masonry acceptance criteria, the original

acceptance criteria used for this condition are the National

Concrete Masonry Associations (NCMA) ``Specification for the Design

and Construction of Load Bearing Masonry--1979'' allowable stresses

times a 1.67 factor. These allowable stresses correspond to stress

equal to the modulus of rupture (fr) of the masonry

divided by a factor of safety of 3.35. During reviews to address

masonry wall issues per NRC IE Bulletin 80-11, six walls did not

meet this acceptance criteria. The acceptance criteria used for

these walls was for fr values determined from testing at

Clinton Power Station divided by a factor of safety of 2.5. This

acceptance criteria was accepted by the NRC for LaSalle in

Supplement 5 of NUREG 0519, Safety Evaluation Report related to the

Operation of LaSalle County Station, Units 1 and 2. The VR exhaust

plenum walls will use the same acceptance criteria for the transient

HELB pressurization cases.

The minimum masonry safety factor for the LaSalle Unit 2 walls

affected by the HELB loads range from 2.6 to 3.1 with one wall

having a safety factor of 4.9.

Masonry wall steel support members were originally designed for

this condition elastically to the American Institute of Steel

Construction's (AISC) ``Steel Construction Manual--Seventh Edition''

allowable stresses times a 1.6 factor. In the reassessment of these

members due to the transient HELB pressurization, elasto-plastic

behavior is allowed (with a ductility ratio limit of 10). It is

appropriate to consider them similar to high-energy line break

systems that will maintain their integrity as they absorb the energy

of the incidental pressure excursion.

High-energy line breaks are discussed in Section 3.6 of the

UFSAR. The discussion in this section focuses on the design of pipe

whip restraints, and in Table 3.6-6 acceptance criteria are

provided. This table shows that the energy absorbing portions of the

pipe whip restraint are allowed to go plastic, thereby absorbing

energy. While Table 3.6-6 of the UFSAR deals with energy absorbing

portions of the pipe whip restraints, wide-flange shapes are not

addressed. Wide-flange shapes absorb energy through flexural

deformations.

Guidance on appropriate acceptance criteria for flexural members

is provided in Appendix A to SRP 3.5.3, ``Barrier Design

Procedures.'' This appendix indicates that for tension due to

flexure in structural steel members, a ductility ratio value not to

exceed 10.0 is acceptable. SRP 3.8.4, paragraph III.5 also notes

that some localized points on the structure, the allowable stresses

specified for ``structural steel'' may be exceeded, provided that

integrity of the structure is not affected.

Note that only one of the Unit 2 walls affected by these HELB

loads required the use of the elasto-plastic acceptance criteria for

two structural steel members.

In summary, these alternate criteria for reassessment of the

integrity of the LaSalle Reactor Building Ventilation Exhaust Plenum

masonry walls in conjunction with the design changes adding a

pressure relief damper, an excess flow check damper and masonry wall

support steel changes, assures that the walls will maintain their

integrity during a MSLB. The safety factor is reduced; however, the

walls have sufficient strength and safety margin to maintain

structural integrity and thus perform their intended safety function

during the pressurization transient due to a MSLB accident.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois

[[Page 32288]]

Valley Community College, Oglesby, Illinois 61348-9692.

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 Section Chief: Anthony J. Mendiola.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant,

Charlevoix, County, Michigan

Date of amendment request: May 11, 1999 (Accession No. 9905170189).

Description of amendment request: The proposed amendment would

delete from the Defueled Technical Specifications (DTS) the definition

for site boundary and Figure 5.1-1, Big Rock Point Site Map, and revise

the description of the Big Rock Point site under subsection 5.1. The

amendment also proposes editorial changes associated with the above

proposed revisions.

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 50.91, Consumers Energy Company has

made a determination that the proposed amendment does not involve

significant hazards considerations. Consumers Energy Company has

concluded that the proposed amendment will 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 proposed change is administrative in nature and has no

[e]ffect on the health and safety of the public. There is no

reduction or elimination of federal regulatory requirements

associated with the proposed amendment. The information being

removed from the Defueled Technical Specifications is unnecessary

since Site Boundary is already defined in 10 CFR Part 20, and the

site map [Defueled Technical Specification Figure 5.1-1] is already

provided in the Updated Final Hazards [Summary] Report. Furthermore,

the proposed changes are consistent with the guidance provide in

NUREG-1625 ['Proposed Standard Technical Specifications for

Permanently Defueled Westinghouse Plants''].

The proposed change does not:

(1) Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed amendment does not change the site boundary as it

currently exists. Deleting the Site Boundary definition and changing

the upper case characters to lower case throughout the DTS and the

Bases where it appears, and deleting the site figure from the DTS

and related references will not increase the probability or

consequences of a new or different kind of accident previously

evaluated. This proposed change is administrative in nature and does

not involve fuel handling or affect or modify any system, structure

or component.

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

accident from any accident previously evaluated.

The proposed amendment does not change the site boundary as it

currently exists. Deleting the Site Boundary definition and changing

the upper case characters to lower case throughout the DTS and the

Bases where it appears, and deleting the site figure from the DTS

and related references will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This proposed change is administrative in nature and does not

involve fuel handling or affect or modify any system, structure or

component.

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

The proposed changes do not involve any physical changes to the

plant or plant procedures. There will be no reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: North Central Michigan

College, 1515 Howard Street, Petosky, MI 49770.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Section Chief: Dr. Michael T. Masnik.

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

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

Date of amendment request: October 2, 1998, supplemented May 13,

1999.

Description of amendment request: The proposed amendments would

resolve an unreviewed safety question involving use of credit for

reactor building overpressure in the licensing basis for the available

net positive suction head for the reactor building spray pumps and the

low pressure injection pumps. If approved, the appropriate changes

would be incorporated in the Oconee Updated Final Safety Analysis

Report.

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.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The reactor building spray (RBS) and low pressure injection

(LPI) systems are not considered as initiators of any analyzed

event, therefore, this change has no impact on the probability of an

event previously analyzed.

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 reactor

building overpressure to be credited in the calculation of available

net positive suction head (NPSH) for the RBS and LPI pumps for a

limited period of time during the sump recirculation phase. It is

supported by calculations which demonstrate that adequate reactor

building overpressure will be available to ensure the RBS and LPI

systems will be capable of performing their safety functions. Thus,

the proposed change does not significantly increase the consequences

of an accident previously evaluated.

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

from the accidents previously evaluated?

The proposed change permits limited reactor building

overpressure to be credited in the calculation of available NPSH for

the RBS and LPI pumps for a limited period of time during the sump

recirculation phase. It does not involve a physical alteration of

the plant. The proposed change is supported by calculations which

demonstrate that adequate reactor building overpressure will be

available to ensure the RBS and LPI systems will be capable of

performing their safety functions. This change will not alter the

manner in which the RBS or LPI system is initiated, nor will the

function demands on the RBS or LPI system be changed. Thus, the

proposed change does not create the possibility of a new or

different kind of accident.

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

The proposed change permits limited reactor building

overpressure to be credited in the calculation of available NPSH for

the RBS and LPI pumps for a limited period of time during the sump

recirculation phase. Crediting a slight amount of overpressure does

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

because conservative analyses demonstrate that adequate reactor

building overpressure will be available to ensure the RBS and LPI

systems will be capable of performing their safety functions. Thus,

the proposed change does not involve a significant reduction in a

margin of safety.

Duke has concluded based on the above information that there are

no significant hazards involved in this LAR.

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

this review, it appears that the three

[[Page 32289]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

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

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

Date of amendment request: May 11, 1999.

Description of amendment request: The proposed amendments would:

(a) revise the pressure-temperature (P-T) limits of Technical

Specification (TS) 3.4.3 for heatup, cooldown, and inservice test

limitations for the Reactor Coolant System to a maximum of 33 Effective

Full Power Years; (b) revise TS 3.4.12, Low Pressure Overpressure

Protection System (LTOP), to reflect the revised P-T limits of the Unit

1, 2, and 3 reactor vessels; (c) permit operation during LTOP

conditions with two reactor coolant pumps in operation in a single

loop; and (d) relax the LTOP operating envelope, thereby reducing

potential challenges to the reactor coolant system power operated

relief valves.

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.

A. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

No.

These proposed Technical Specification (TS) changes were

developed utilizing the procedures of ASME XI, Appendix G, in

conjunction with Code Cases N-514, N-588 and N-626, as described in

the Technical Justification. Usage of these procedures provides

compliance with the underlying intent of 10 CFR 50 Appendix G and

provide safety limits and margins of safety that ensure failure of a

reactor vessel will not occur.

The proposed changes do not impact the capability of the reactor

coolant pressure boundary (i.e., no change in operating pressure,

materials, seismic loading, etc.) and therefore do not increase the

potential for the occurrence of a loss of coolant accident (LOCA).

The changes do not modify the reactor coolant system pressure

boundary, nor make any physical changes to the facility design,

material, or construction standards. The probability of any design

basis accident (DBA) is not affected by this change, nor are the

consequences of any DBA affected by this change. The proposed

Pressure-Temperature (P-T) limits, Low Temperature Overpressure

(LTOP) limits and setpoints, and allowable operating reactor coolant

pump combinations are not considered to be an initiator or

contributor to any accident analysis addressed in the Oconee UFSAR.

The proposed changes do not adversely affect the integrity of

the RCS such that its function in the control of radiological

consequences is affected. Radiological off-site exposures from

normal operation and operational transients, and faults of moderate

frequency do not exceed the guidelines of 10 CFR 100. In addition,

the proposed changes do not affect any fission product barrier. The

revised PORV LTOP setpoint is established to protect reactor coolant

pressure boundary. The changes do not degrade or prevent the

response of the PORV or safety-related systems to previously

evaluated accidents. In addition, the changes do not alter any

assumption previously made in the mitigation of the radiological

consequences of an accident previously evaluated.

Therefore, the probability or consequences of an accident

previously evaluated will not be increased by approval of the

requested changes.

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

from the accident previously evaluated?

No.

The proposed license amendment revises the Oconee reactor vessel

P-T limits, LTOP limits and setpoints, and allowable operating

reactor coolant pumps combinations. Compliance with 10 CFR 50

Appendix G, includes utilization of ASME XI, Appendix G, as modified

by Code Cases N-514, N-588 and N-626 to meet the underlying intent

of the regulations.

Operation of Oconee in accordance with these proposed Technical

Specifications changes will not create any failure modes not bounded

by previously evaluated accidents. Consequently, approval of these

changes will not create the possibility of a new or different

accident from any accident previously evaluated.

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

No.

The proposed Technical Specification (TS) changes were developed

utilizing the procedures of ASME XI, Appendix G, in conjunction with

Code Cases N-514, N-588 and N-626, as described in the Technical

Justification. Usage of these procedures provides compliance with

the underlying intent of 10 CFR 50 Appendix G and provides safety

limits and margins of safety which ensure failure of a reactor

vessel will not occur.

No plant safety limits, set points, or design parameters are

adversely affected. The fuel, fuel cladding, and Reactor Coolant

System are not impacted. Therefore, there will be no significant

reduction in any margin of safety as a result of approval of the

requested changes.

Duke has concluded based on this information there are no

significant hazards considerations involved in this amendment

request.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

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

Generating Station, Salem County, New Jersey

Date of amendment request: May 17, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications associated with the enabling of the

Oscillation Power Range Monitor (OPRM) instrumentation reactor

protection system (RPS) trip function. The OPRM is designed to detect

the onset of reactor core power oscillations resulting from thermal-

hydraulic instability and suppresses them by initiating a reactor scram

via the RPS trip logic.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed change specifies limiting conditions for

operations, required actions and surveillance requirements of the

OPRM system and allows operation in regions of the power to flow map

currently restricted by the requirements of Interim Corrective

Actions (ICAs) and certain limiting conditions of operation of

Technical Specifications (TS) 3.4.1. The OPRM system can

automatically detect and suppress conditions necessary for thermal-

hydraulic (T-H) instability. A T-H instability event has the

potential to challenge the Minimum Critical Power (MCPR) safety

limit. The restrictions of the ICAs and TS 3.4.1 were imposed to

ensure adequate capability to detect and suppress conditions

consistent with the onset of T-H oscillations that may develop into

a T-H instability event. With the installation of the OPRM System,

these restrictions are no longer required.

[[Page 32290]]

The probability of a T-H instability event is most significantly

impacted by power to flow conditions such that only during operation

inside specific regions of the power to flow map, in combination

with power shape and inlet enthalpy conditions, can the occurrence

of an instability event be postulated to occur. Operation in these

regions may increase the probability that operation with conditions

necessary for a T-H instability can occur.

However, when the OPRM is operable with operating limits as

specified in the COLR [Core Operating Limits Report], the OPRM can

automatically detect the imminent onset of local power oscillations

and generate a trip signal. Actuation of an RPS trip will suppress

conditions necessary for T-H instability and decrease the

probability of a T-H instability event. In the event the trip

capability of the OPRM is not maintained, the proposed change

includes actions which limit the period of time before the effected

OPRM channel (or RPS system) must be placed in the trip condition.

If these actions would result in a trip function, an alternate

method to detect and suppress thermal hydraulic oscillations is

required. In either case the duration of this period of time is

limited such that the increase in the probability of a T-H

instability event is not significant. Therefore the proposed change

does not result in a significant increase in the probability of an

accident previously evaluated.

An unmitigated T-H instability event is postulated to cause a

violation of the MCPR safety limit. The proposed change ensures

mitigation of T-H instability events prior to challenging the MCPR

safety limit if initiated from anticipated conditions by detection

of the onset of oscillations and actuation of an RPS trip signal.

The OPRM also provides the capability of an RPS trip being generated

for T-H instability events initiated from unanticipated but

postulated conditions. These mitigating capabilities of the OPRM

system would become available as a result of the proposed change and

have the potential to reduce the consequences of anticipated and

postulated T-H instability events. Therefore, the proposed change

does not significantly increase the 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 specifies limiting conditions for

operations, required actions and surveillance requirements of the

OPRM system and allows operation in regions of the power to flow map

currently restricted by the requirements of ICAs and TS 3.4.1. The

OPRM system uses input signals shared with APRM [Average Power Range

Monitor] and rod block functions to monitor core conditions and

generate an RPS trip when required. Quality requirements for

software design, testing, implementation and module self-testing of

the OPRM system provide assurance that no new equipment malfunctions

due to software errors are created. The design of the OPRM system

also ensures that neither operation nor malfunction of the OPRM

system will adversely impact the operation of other systems and no

accident or equipment malfunction of these other systems could cause

the OPRM system to malfunction or cause a different kind of

accident. Therefore, operation with the OPRM system does not create

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

accident previously evaluated.

Operation in regions currently restricted by the requirements of

ICAs and TS 3.4.1 is within the nominal operating domain and ranges

of plant systems and components for which postulated equipment and

accidents have been evaluated. Therefore operation within these

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

of accident from any accident previously evaluated.

The proposed change which specifies limiting conditions for

operations, required actions and surveillance requirements of the

OPRM system and allows operation in certain regions of the power to

flow [map] 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 specifies limiting conditions for

operations, required actions and surveillance requirements of the

OPRM system and allows operation in regions of the power to flow map

currently restricted by the requirements of ICAs and TS 3.4.1.

The OPRM system monitors small groups of LPRM signals for

indication of local variations of core power consistent with T-H

oscillations and generates an RPS trip when conditions consistent

with the onset of oscillations are detected. An unmitigated T-H

instability event has the potential to result in a challenge to the

MCPR safety limit. The OPRM system provides the capability to

automatically detect and suppress conditions which might result in a

T-H instability event and thereby maintains the margin of safety by

providing automatic protection for the MCPR safety limit while

significantly reducing the burden on the control room operators. In

the event the trip capability of the OPRM is not maintained, the

proposed change includes actions which limit the period of time

before the effected OPRM channel (or RPS system) must be placed in

the trip condition. If these actions would result in a trip

function, an alternate method to detect and suppress thermal

hydraulic oscillations is required. Since, in either case, the

duration of this period of time is limited so that the increase in

the probability of a T-H instability event is not significant.

Operation with the OPRM system does not involve a significant

reduction in a margin of safety.

Operation in regions currently restricted by the requirements of

ICAs and TS 3.4.1 is within the nominal operating domain assumed for

identifying the range of initial conditions considered in the

analysis of anticipated operational occurrences and postulated

accidents. Therefore, operation in these regions does not involve a

significant reduction in the margin of safety.

The proposed change, which specifies limiting conditions for

operations, required actions and surveillance requirements of the

OPRM system and allows operation in certain regions of the power to

flow map, does not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

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

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

NRC Section Chief: James W. Clifford.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station

(VCSNS), Unit No. 1, Fairfield County, South Carolina

Date of amendment request: May 17, 1999.

Description of amendment request: The proposed amendment would

change VCSNS Technical Specification 3.7.1.3 ``Condensate Storage

Tank--Limiting Conditions for Operation'' to revise the tank minimum

contained water volume from 172,000 gallons to 179,850 gallons.

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. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

FSAR [Final Safety Analysis Report] 10.4.9.1 states that minimum

required usable volume for the Condensate Storage Tank (CST) is

158,570 gallons based on maintaining the plant at HOT STANDBY

conditions for eleven hours. This volume has already been adjusted

for both plant uprate conditions and replacement steam generator

requirements. This change to LCO [Limiting Condition for Operation]

3.7.1.3 will ensure that 160,054 gallons is maintained in the CST,

being available and dedicated to the Emergency Feedwater (EFW)

System. Thus, this change will ensure that the EFW System has an

adequate water supply to perform its design basis function in regard

to maintaining the plant in HOT STANDBY condition.

2. This request does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This change increases the minimum required volume of water in

the CST, thus

[[Page 32291]]

ensuring that the EFW System can perform its required safety

function. The maximum and normal water levels in the CST are not

being changed. Therefore, no new failure modes of the CST, or

flooding concerns are created.

3. This request does not involve a significant reduction in a

margin to safety[.]

This change does not reduce any margin associated with the CST

inventory available to the EFW. In fact, a small gain in margin

(less than 1%) is realized by specifying the minimum required volume

based on the maximum volume available due to nozzle locations and

other physical characteristics of the tank instead of the minimum

required to maintain HOT STANDBY for 11 hours. Additionally, the

requirement for sufficient CST volume to maintain HOT STANDBY for 11

hours is still met and the Service Water System still provides the

long term supply of safety grade cooling water to the EFW System.

The Service Water supply is not affected by this change, and thus

the margin for safety grade cooling water to the EFW System (or

safety grade cooling of the RCS) is not affected.

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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Section Chief: Richard L. Emch, Jr.

Southern Nuclear Operating Company, Inc, Docket No. 50-348 Joseph M.

Farley Nuclear Plant Unit 1, Houston County, Alabama

Date of amendment request: April 30, 1999.

Description of amendment request: The proposed amendment would add

an additional condition to the Farley Nuclear Plant (FNP), Unit 1

license. This condition would allow cycle 16 operation based on a risk-

informed approach to evaluate steam generator tube structural

integrity.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the FSAR [Final Safety Analysis Report]. The probability of tube

burst is slightly increased as a result of this proposed amendment

but is within current industry guidance. Therefore, the probability

of a previously evaluated accident are not significantly increased.

There is no change in the FNP design basis as a result of this

change and, as a result, this change does not involve a significant

increase in the consequences of an accident previously evaluated.

The proposed changes to the TSs [technical specifications] do

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

than any accident already evaluated in the FSAR. No new limiting

single failure or accident scenario has been created or identified

due to the proposed changes. Safety-related systems will continue to

perform as designed. The proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

The proposed changes do not involve a significant reduction in

the margin of safety. There is no impact in the accident analyses.

These proposed changes are technically consistent with the

requirements of NEI [Nuclear Energy Institute] 97-06, ``Steam

Generator Program Guidelines,'' Draft Regulatory Guide DG 1074,

``Steam Generator Tube Integrity,'' and Regulatory Guide (RG) 1.174,

``An Approach for Using Probabilistic Risk Assessment In Risk-

Informed Decisions on Plant-Specific Changes to the Licensing

Basis.'' Thus the proposed changes do not involve a significant

reduction in the margin of safety.

Accordingly, SNC [Southern Nuclear Operating Company] has

determined that the proposed amendment to the Facility Operating

License NPF-2 does not involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

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

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama.

NRC Section Chief: Richard L. Emch, Jr.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

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

Virginia

Date of amendment request: May 3, 1999.

Description of amendment request: The proposed changes will modify

the Technical Specifications to ensure the emergency ventilation system

is maintained operable consistent with the assumptions in the

radiological dose consequence reanalysis from a Large Break Loss-of-

Coolant Accident and to clearly identify that the ventilation system is

a shared system between the two units.

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. There is no significant change in the probability or

consequences of an accident previously evaluated. There are no

system changes which would increase the probability of occurrence of

an accident. The dose consequences of the accidents have been

reviewed, and in some cases the doses at the EAB [exclusion area

boundary] * * * and the doses to the control room personnel were

found to increase. However, this increase is not significant because

the revised doses remain below the limits of 10 CFR 100 and below

the limits of GDC [General Design Criterion]--19 of Appendix A of 10

CFR 50.

2. No new accident types or equipment malfunction scenarios have

been introduced. Therefore, the possibility of an accident of a

different type than any evaluated previously in the UFSAR [Updated

Final Safety Analysis Report] is not created.

3. There is no significant reduction in the margin of safety, as

the revised dose calculations for all accidents continue to meet the

appropriate GDC-19 limits.

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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Section Chief: Richard L. Emch, Jr.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

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

Virginia

Date of amendment request: May 6, 1999.

Description of amendment request: The proposed changes will modify

the Technical Specifications, revising the surveillance frequency for

the Reactor Trip System (RTS) and Engineered Safety Features Actuation

System (ESFAS) analog instrumentation

[[Page 32292]]

channels and also revising the allowed outage time and action times for

the RTS and ESFAS analog instrumentation channels and the actuation

logic.

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:

Virginia Electric and Power Company has reviewed the

requirements of 10 CFR 50.92 as they relate to the proposed Reactor

Trip System (RTS) and Engineered Safety Features Actuation System

(ESFAS) Technical Specification changes for the North Anna Units 1

and 2 and determined that a significant hazards consideration is not

involved. In support of this conclusion, the following evaluation is

provided.

Criterion 1--Operation of North Anna Units 1 and 2 in accordance

with the proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The determination that the results of the

proposed changes remain within acceptable criteria was established

in the SER(s) [Safety Evaluation Reports] prepared for WCAP-10271,

WCAP-10271 Supplement 1, WCAP-10271 Supplement 2, WCAP-10271

Supplement 2, Revision 1 and WCAP-14333 issued by letters dated

February 21, 1985, February 22, 1989, April 30, 1998, and July 15,

1998.

Implementation of the proposed changes is expected to result in

an increase in total RTS and ESFAS yearly unavailability. The

proposed changes have been shown to result in a small increase in

the core damage frequency (CDF) due to the combined effects of

increased RTS and ESFAS unavailability and reduced inadvertent

reactor trips.

The values determined by the WOG [Westinghouse Owners Group] and

presented in the WCAP for the increase in CDF were verified by

Brookhaven National Laboratory (BNL) as part of an audit and

sensitivity analyses for the NRC [Nuclear Regulatory Commission]

Staff. Based on the small value of the increase compared to the

range of uncertainty in the CDF, the increase is considered

acceptable. The analysis performed by the WOG and presented in the

WCAP included changes to the surveillance frequencies for the

automatic actuation logic and actuation relays and the reactor trip

and bypass breakers. The overall increase in the CDF, including the

changes to the surveillance frequencies for the automatic actuation

logic and actuation relays and the reactor trip and bypass breakers,

was approximately 6 percent. However, even with this increase, the

overall CDF remains lower than the NRC safety goal of

10-4/reactor year.

Changes to surveillance test frequencies for the RTS and ESFAS

interlocks do not represent a significant reduction in testing. The

currently specified test interval for interlock channels allows the

surveillance requirement to be satisfied by verifying that the

permissive logic is in its required state using the annunciator

status light. The surveillance as currently required only verifies

the status of the permissive logic and does not address verification

of channel setpoint or operability. The setpoint verification and

channel operability is verified after a refueling shutdown. The

definition of the channel check includes comparison of the channel

status with other channels for the same parameter. The requirement

to routinely verify permissive status is a different consideration

than the availability of trip or actuation channels which are

required to change state on the occurrence of an event and for which

the function availability is more dependent on the surveillance

interval. Therefore, the change in the interlock surveillance

requirement to at least once every 18 months does not represent a

significant change in channel surveillance and does not involve a

significant increase in unavailability of the RTS and ESFAS.

For the additional relaxations in WCAP-14333, the WOG evaluated

the impact of the additional relaxation of allowed outage times and

completion times, and action statements on core damage frequency.

The change in core damage frequency is 3.1 percent for those plants

with two out of three logic schemes that have not implemented the

proposed surveillance test interval, allowed outage times, and

completion times evaluated in WCAP-10271 and its supplements. This

analysis calculates a significantly lower increase in core damage

frequency than the WCAP-10271 analysis calculated. This can be

attributed to more realistic maintenance intervals used in the

current analysis and crediting the AMSAC [ATWS (anticipated

transient without scram) mitigating system actuation circuitry]

system as an alternative method of initiating the auxiliary

feedwater pumps. Therefore, the overall increase in CDF is estimated

to be 3.1% for the proposed changes per the generic Westinghouse

analysis.

The NRC performed an independent evaluation of the impact on

core damage frequency (CDF) and large early release fraction (LERF).

The results of the staff's review indicate that the increase in core

damage frequency is small (approximately 3.2%) and the large early

release fraction would increase by only 4 percent for 2 out of 3

logic schemes that have not implemented the proposed surveillance

test interval, allowed outage times, and completion times evaluated

in WCAP-10271 and its supplements. Further, the absolute values for

CDF still remain within NRC safety goals.

Therefore, the proposed changes do not result in a significant

increase in the severity or consequences of an accident previously

evaluated. Implementation of the proposed changes affects the

probability of failure of the RTS and ESFAS but does not alter the

manner in which protection is afforded or the manner in which

limiting criteria are established.

Criterion 2--The proposed license amendment does not create the

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

previously evaluated.

The proposed changes do not result in a change in the manner in

which the RTS or ESFAS provide plant protection. No change is being

made which alters the functioning of the RTS or ESFAS (other than in

a test mode). Rather the likelihood or probability of the RTS or

ESFAS functioning properly is affected as described above.

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

new or different kind of accident as defined in the Safety Analysis

Report.

The proposed changes do not involve hardware changes. Some

existing instrumentation is designed to be tested in bypass and

current Technical Specifications allow testing in bypass. Testing in

bypass is also recognized by IEEE [Institute of Electrical and

Electronics Engineers] Standards. Therefore, testing in bypass has

been previously approved and implementation of the proposed changes

for testing in bypass does not create the possibility of a new or

different kind of accident from any previously evaluated.

Furthermore since the other proposed changes do not alter the

physical operation or functioning of the RTS or ESFAS the

possibility of a new or different kind of accident from any

previously evaluated has not been created.

Criterion 3--The proposed license amendment does not involve a

significant reduction in a margin of safety.

The proposed changes do not alter the safety limits, limiting

safety system setpoints or limiting conditions for operation. The

RTS and ESFAS analog instrumentation remain operable to mitigate as

assumed in the accident analysis. The impact of reduced testing

other than as addressed above is to allow a longer time interval

over which instrument uncertainties (e.g., drift) may act.

Implementation of the proposed changes is expected to result in

an overall improvement in safety by less frequent testing of the RTS

and ESFAS analog instruments will result in less inadvertent reactor

trips and actuation of Engineered Safety Features components.

This analysis demonstrates that the proposed amendment to The

North Anna Unit 1 and 2 Technical Specifications does not involve a

significant increase in the probability or consequences of a

previously evaluated accident, does not create the possibility of a

new or different kind of accident and 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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

[[Page 32293]]

NRC Section Chief: Richard L. Emch Jr.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

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

County, Illinois

Date of application for amendments: March 22, 1999.

Brief description of amendments: The amendments modify the

technical specifications to permit the use of the Gamma-Metrics Post

Accident Neutron Monitors source range neutron flux detectors in

addition to the Westinghouse source range neutron flux monitors to

satisfy the requirement that two source range neutron flux monitors be

operable during Mode 6 operations (refueling).

Date of issuance: June 2, 1999.

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

Amendment Nos.: 109 & 109, 102 & 102.

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 29, 1999 (64 FR

14944). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 2, 1999.

No significant hazards consideration comments received: No.

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

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

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

Wilmington, Illinois 60481.

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

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

Date of application for amendments: December 2, 1996, as

supplemented on May 27, 1999.

Brief description of amendments: The amendments revised Technical

Specification 3/4.4.2 to reduce the number of required Safety/Relief

valves (SRVs). This change supports a modification to remove five of

the currently installed SRVs due to excess capacity and to reduce the

amount of valve maintenance and associated worker radiation dose. The

revised TS requires that 12 of the remaining installed 13 SRVs be

operable.

Date of issuance: June 3, 1999.

Effective date: Immediately, to be implemented prior to startup of

L1C10 for Unit 1 and prior to startup of L2C9 for Unit 2.

Amendment Nos.: 133 & 118.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4343). The May 27, 1999, submittal 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 June 3, 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.

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

Michigan

Date of application for amendment: March 23, 1999 (NRC-99-0025).

Brief description of amendment: The amendment revises Technical

Specification Surveillance Requirement (SR) 4.4.1.1.1 to require each

recirculation pump discharge valve be demonstrated operable at least

once every 18 months, deletes the ``*'' footnote from the SR, and

revises the footnote itself to read ``Not used.''

Date of issuance: May 25, 1999.

Effective date: May 25, 1999, with full implementation within 90

days.

Amendment No.: 133.

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

Technical Specifications.

Date of initial notice in Federal Register: April 21, 1999 (64 FR

19555)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

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

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

Date of application for amendments: July 9, 1998, as supplemented

March 31, 1999.

Brief description of amendments: These amendments revised Technical

Specification (TS) 3/4.7.1.1 and associated Bases for both units. This

amendment specifies maximum allowable reactor power level based on the

number of operable main steam safety valves (MSSVs) rather than

requiring reduction in reactor trip setpoint. This change is consistent

with the Nuclear Regulatory Commission's improved Standard Technical

Specifications for Westinghouse plants (NUREG-1431, Revision 1). The

maximum allowable reactor power level with inoperable MSSVs will be

calculated based on the recommendations of Westinghouse Nuclear Safety

Advisory Letter 94-01. The change to the Unit 1 TS 3.7.1.1 also deletes

reference to 2 loop operation since 2 loop operation is not a licensed

[[Page 32294]]

condition for either unit. Unit 1 TS Table 3.7-3 is then renumbered to

be Table 3.7-2.

The March, 31, 1999 letter withdrew a portion of the amendment

which would have removed the values of the orifice diameter of each

MSSV from the TSs. This information will be maintained in the TSs.

Date of issuance: June 3, 1999.

Effective date: Units 1 and 2 as of date of issuance and shall be

implemented within 60 days.

Amendment Nos.: 223 and 99.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43203). The March 31, 1999 letter did not change the initial proposed

no significant hazards consideration determination or expand the

amendment beyond the scope of the initial notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 3, 1999.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: August 31, 1998.

Brief description of amendment: Changes the Crystal River Unit 3

Technical Specifications to add additional instrumentation variables to

Improved Technical Specification Table 3.3.17-1, Post-Accident

Monitoring Instrumentation.

Date of issuance: June 3, 1999.

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

commencing cycle 12 operation.

Amendment No.: 177.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

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

56250).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

Nuclear Generating Plant, Unit 3, Citrus County, Florida

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

supplemented January 29 and May 7, 1999.

Brief description of amendment: The amendment changes the Improved

Technical Specifications for several reactor protection system and

engineered safeguards actuation system setpoint values, and changes the

surveillance requirement to verify valve position for valves in the

high pressure injection system flowpath.

Date of issuance: May 21, 1999.

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

commencing Cycle 12 operation.

Amendment No.: 178.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

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

FR 71966). The supplemental letters dated January 29 and May 7, 1999,

did not change the original proposed no significant hazards

consideration determination, or expand the scope of the amendment

request as originally noticed.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

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

Date of amendment request: December 16, 1998.

Description of amendment request: These amendments consist of

changes to the Technical Specifications (TS) in response to Florida

Power & Light's (FPL) application dated December 16, 1998, regarding

facility staff qualifications for multi-discipline supervisor (MDS)

positions at Lucie Units 1 and 2. The amendments revise the

administrative controls in TS Section 6.3, ``Unit Staff

Qualifications,'' by modifying FPL's commitment to ANSI/ANS 3.1-1978,

``Selection and Training of Nuclear Power Plant Personnel,'' to

incorporate specific staff qualifications for the position of MDS.

Date of Issuance: May 25, 1999.

Effective Date: May 25, 1999.

Amendment Nos.: 161 and 102.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the TS.

Date of Initial Notice in Federal Register: February 10, 1999 (64

FR 6698).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

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

Generating Station, Ocean County, New Jersey

Date of application for amendment: November 5, 1998, as

supplemented February 18, 1999.

Brief description of amendment: The amendment modifies the safety

limits and surveillances of the LPRM and APRM systems and related Bases

pages to ensure the APRM channels respond within the necessary range

and accuracy and to verify channel operability. In addition, an

unrelated change to the Bases of Specification 2.3 is included to

clarify some ambiguous language.

Date of Issuance: June 2, 1999.

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

within 30 days.

Amendment No.: 208.

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

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

FR 69342). The February 18, 1999, supplemental letter provided

clarifying information, was within the scope of the original

application, and did not change the staff's original no significant

hazards consideration determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated June 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336,

and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New

London County, Connecticut

Date of application for amendment: December 22, 1998, as

supplemented March 19, 1999.

Brief description of amendment: The amendment replaces specific

titles in Section 6.0 of the Technical

[[Page 32295]]

Specifications of all three Millstone units with generic titles.

Date of issuance: June 3, 1999.

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

30 days from the date of issuance.

Amendment No.: 105, 235, and 171.

Facility Operating License Nos. DPR-21, DPR-65, and NPF-49:

Amendment revised the Technical Specifications.

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

4158). The March 19, 1999 letter provided clarifying information that

did not change the scope of the December 22, 1998, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 3, 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.

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 application for amendments: April 20, 1999.

Brief description of amendments: The amendments revised the

implementation date for the relocation of the requirements specified in

Technical Specification Sections 3.1.E and 5.1 to the Updated Final

Safety Analyis Report. On December 7, 1998, the NRC had previously

issued license amendments 141 and 132 for Units 1 and 2, respectively,

approving the relocation of aforementioned requirements by June 1,

1999. The proposed amendments would postpone the implementation date to

September 1, 1999.

Date of issuance: June 2, 1999.

Effective date: June 2, 1999, with full implementation within 30

days .

Amendment Nos.: 145 and 136.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 29, 1999 (64 FR

23131) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendments: January 4, 1999.

Brief description of amendments: These amendments revise the

administrative section of the Technical Specification pertaining to

controlled access to high radiation areas, and the reporting dates for

the annual occupational radiation exposure report and the annual

radioactive effluent release report.

Date of issuance: May 24, 1999.

Effective date: Units 1 and 2, as of date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 135 and 100.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

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

FR 6706) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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: January 25, 1999.

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) by relocating certain requirements from the TSs to

the Final Safety Analysis Report.

Date of issuance: May 24, 1999.

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

30 days.

Amendment No.: 189.

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

Technical Specifications.

Date of initial notice in Federal Register: April 21, 1999 (64 FR

19562).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit

1, Luzerne County, Pennsylvania

Date of application for amendment: March 12, 1999.

Brief description of amendment: This amendment would change the

allowable values for both the core spray system and the low pressure

coolant injection system reactor steam dome pressure-low functions.

Date of issuance: May 25, 1999.

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

within 30 days after startup from the Unit 1 eleventh refueling and

inspection outage currently scheduled for spring 2000.

Amendment No.: 181.

Facility Operating License No. NPF-14: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 7, 1999 (64 FR

17028).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 25, 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.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment: October 27, 1998, as

supplemented by letters in 1999 dated January 11, January 29, February

25, and April 7 (two letters), and May 17.

Brief description of amendment: The amendment revised Technical

Specification 4.4.5.4, Table 4.4-3 and the associated Bases to allow

the repair of the steam generator tubes with the Electrosleeve tube

repair method.

Date of issuance: May 21, 1999.

Effective date: May 21, 1999, to be implemented within 30 days from

the date of issuance. The amendment includes a two cycle operating

limit that requires all steam generator tubes repaired with

Electrosleeves to be removed from service at the end of two operating

cycles following installation of the first Electrosleeve in the steam

generators.

Amendment No.: 132.

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

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

66604). The supplemental letters in 1999 dated January 11, January 29,

February 25, and April 7 (two letters)

[[Page 32296]]

provided additional clarifying information that did not expand the

staff's original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Elmer Ellis Library,

University of Missouri, Columbia Missouri 65201.

Dated at Rockville, Maryland, this 9th day of June 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-15098 Filed 6-15-99; 8:45 am]

BILLING CODE 7590-01-P