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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission

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

notice. Pub. L. 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 October 24, 1998, through November 5, 1998.

The last biweekly notice was published on November 4, 1998 (63 FR

59584).

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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 December 18, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

be filed with 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

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Public Document Room, the Gelman Building, 2120 L Street, NW.,

Washington DC, by the above date. A copy of the petition should also be

sent to the Office of the General Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, and to the attorney for the

licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: October 16, 1998.

Description of amendment request: The proposed amendments would

lower the power level below which the turbine control valve (TCV) and

turbine stop valve (TSV) closure scram signals and the end-of-cycle

recirculation pump trip (EOC-RPT) signals are not in effect. The bypass

setpoint (Pbypass) would be reduced from 30 percent rated

power to 25 percent rated power. The licensee also proposes to delete

the reference to turbine first stage pressure as a measure of core

thermal power in the Technical Specifications. To ensure that the trip

functions will not be inadvertently bypassed when they are required to

be operable, a requirement would be added to periodically verify that

TCV and TSV scram trip functions and the ECO-RPT trip functions are not

bypassed at greater than or equal to 25 percent of rated thermal power.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated:

The probability of an accident previously evaluated will not

increase as a result of this change because the setpoint change does

not alter any of the initiators of an accident or cause them to

occur more frequently.

The consequences of an accident previously evaluated are not

impacted. LaSalle Units 1 and 2 each have approximately 30 percent

bypass capability. Therefore, a scram on TCV or TSV closure signals

is not needed until 30 percent core thermal power is reached, as

adequate steam bypass capacity is available. A lower

Pbypass remains conservative with respect to this

criterion.

LaSalle utilizes power and flow dependent thermal limits. The

power dependent portion of these thermal limits is dependent on the

Pbypass setpoint. These limits provide assurance that

adequate fuel thermal-mechanical margin is maintained through

adherence to the thermal limits Technical Specification

requirements.

Revised thermal limits have been determined based on the results

of GE transient analyses. Adhering to these thermal limits ensures

that the consequences of an accident or transient would not be

increased from the consequences under the approved 30 percent

setpoint. Adjustments to the thermal limits were determined through

use of the NRC-approved ODYN reactor dynamic model for the limiting

Load Rejection Without Bypass and the Feedwater Controller Failure

events.

The deletion of the reference to turbine first stage pressure

and rewording the Technical Specifications Notes does not affect

either accident initiators or plant equipment, as they are

administrative changes.

Adding the periodic verification that the bypass channels are

set correctly ensures that scrams or EOC-RPT will not be

inadvertently bypassed when Thermal Power is greater than or equal

to 25 percent of Rated Thermal Power. The statement that

specification 4.0.2 applies to the 18 month interval is needed,

since the notes are not standard surveillance requirements and the

interval is consistent with other similar instrumentation to which

4.0.2 currently applies.

Therefore, the proposed changes do not involve a significant

increase in the probability of occurrence or consequences of an

accident previously evaluated.

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

accident from any accident previously evaluated:

The setpoint change and proposed bypass verification notes

ensure that the scrams for TSV closure and TCV fast closure, and

EOC-RPT, will be enabled above 25 percent of rated thermal power,

rather than above 30 percent of rated thermal power. This change

results in simplified reload transient analyses and does not impact

any other equipment.

No other physical modifications are being proposed by this

submittal. The only plant operational impact is that between 25

percent and 30 percent power, the plant will now scram upon a

turbine trip, which is an analyzed transient.

The remaining changes to Technical Specification wording are

administrative in nature and consistent with other Technical

Specifications.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

LaSalle Units 1 and 2 each have approximately 30 percent bypass

capability. Therefore, a scram on TCV or TSV closure signals in not

needed until 30 percent core thermal power is reached, as adequate

steam bypass capacity is available. However, reduction of this

setpoint to 25 percent power actually aids the plant transient

response between 25 percent and 30 percent power.

The new thermal limits reflect the revised setpoint and have

been determined based on revised limiting transient analyses that

have included the new Pbypass value. If a transient were

to occur, the revised operating limits ensure that adequate margin

would be available to preclude violation of the Minimum Critical

Power Ratio (MCPR) safety limit and the fuel thermal-mechanical

limits.

All other UFSAR [Updated Final Safety Analysis Report] events

are either bounded by the analyses performed or are not impacted by

the Pbypass change.

The wording changes to the Technical Specifications do not

change the requirement for the bypass function and for maintaining

the bypass function and thus do not affect the analyses discussed

above.

The addition of the Notes periodically verifying the TCV and TSV

Closure Trip Functions are not bypassed at greater than or equal to

25 percent Rated Thermal Power ensures the trip functions will not

be inadvertently bypassed when required to be Operable.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

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

Illinois 61348-9692.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Stuart A. Richards.

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

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

Date of amendment request: July 22 and October 22, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) to reflect the licensee's

planned use of fuel supplied by Westinghouse. The

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Westinghouse fuel has different design characteristics from the fuel

currently in use. Accordingly, the following changes would need to be

made to the TS: Figure 2.1.1-1, ``Reactor Core Safety Limits--Four

Loops in Operation''; various core operating parameters specified by

Surveillance Requirements 3.2.1.2, 3.2.1.3, and 3.2.2.2; Section 4.2.1,

``Fuel Assemblies''; and Section 5.6.5, ``Core Operating Limits Report

(COLR).''

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, addressing the three standards of 10 CFR 50.92(c):

First Standard

Implementation of this LAR [license amendment request] would not

involve a significant increase in the probability or consequences of

an accident previously evaluated. The revised Reactor Core Safety

Limits Figure further restricts acceptable operation. Moving an

uncertainty factor from the Improved Technical Specifications to the

Core Operating Limits Report (COLR) does not exempt this factor from

regulatory restrictions. COLR parameters are generated by NRC

approved methods with the intent of ensuring that previously

evaluated accidents remain bounding. The COLR is submitted to the

NRC upon implementation of each fuel cycle or when the document is

otherwise revised. No accident probabilities or consequences will be

impacted by this LAR.

Second Standard

Implementation of this LAR would not create the possibility of a

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

revised Reactor Core Safety Limits Figure further restricts

acceptable operation. Moving an uncertainty factor from the Improved

Technical Specifications to the COLR does not exempt this factor

from regulatory restrictions. Since the parameter in question is not

being deleted, the possibility of a new or different kind of

accident from any previously evaluated does not exist.

Third Standard

Implementation of this LAR would not involve a significant

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

confidence in the ability of the fission product barriers to perform

their design functions during and following an accident situation.

These barriers include the fuel cladding, the reactor coolant

system, and the containment system. Use of the ZIRLOTM

cladding material has been reviewed and approved in Reference 1 (as

listed in Chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P,

Duke Power Company Westinghouse Fuel Transition Report).

ZIRLOTM cladding has been extensively used in

Westinghouse nuclear reactors. The changes proposed in this LAR are

necessary to ensure that the performance of the fission product

barriers (cladding) will not be impacted following the replacement

of one fuel design for another. No safety margin will be

significantly impacted.

The NRC staff reviewed the licensee's analysis, and agrees that the

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

staff proposes to determine that the amendment request involves no

significant hazards consideration.

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

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

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

North Carolina.

NRC Project Director: Herbert N. Berkow.

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 amendment request: June 18, 1996. This notice supersedes

the notice published on July 31, 1996 (61 FR 40015) in its entirety.

Description of amendment request: For Beaver Valley Power Station,

Unit No. 1 (BVPS-1) only, the proposed amendment would revise Technical

Specification (TS) 4.4.5 and associated Bases; the Bases for TS 3/

4.4.6.2 would also be revised. The proposed changes are editorial in

nature and are intended to provide consistency between the TSs and

associated Bases. Index page XIX would be revised to reflect the

revision of page numbers for TS Tables 4.4-1 and 4.4-2 due to shifting

of text.

For Beaver Valley Power Station, Unit No. 2 (BVPS-2) only, the

proposed amendment would implement a voltage-based repair criteria for

steam generator tubes similar to the changes approved for BVPS-1 by

License Amendment No. 198. The proposed changes are intended to reflect

the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based

Repair Criteria for Westinghouse Steam Generator Tubes Affected by

Outside Diameter Stress Corrosion Cracking.'' The proposed changes

would revise TSs 4.4.5 and 3.4.6.2 and associated Bases. TS Table 4.4-2

would be revised to reference TS 6.6 for reporting requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Tube burst criteria are inherently satisfied during normal

operating conditions due to the proximity of the tube support plate

(TSP). Test data indicates that tube burst cannot occur within the

TSP, even for tubes which have 100% throughwall electric discharge

machining notches, 0.75 inch long, provided that the TSP is adjacent

to the notched area. Since tube-to-TSP proximity precludes tube

burst during normal operating conditions, use of the criteria must

retain tube integrity characteristics which maintain a margin of

safety of 1.43 times the bounding faulted condition, main steamline

break (MSLB) pressure differential. The Regulatory Guide (RG) 1.121

criterion requiring maintenance of a safety factor of 1.43 times the

MSLB pressure differential on tube burst is satisfied by \7/8\''

diameter tubing with bobbin coil indications with signal amplitudes

less than 8.6 volts, regardless of the indicated depth measurement.

The upper voltage repair limit (VURL) will be

determined prior to each outage using the most recently approved NRC

database to determine the tube structural limit (VSL).

The structural limit is reduced by allowances for nondestructive

examination (NDE) uncertainty (VNDE) and growth

(VGR) to establish VURL. Using the Generic

Letter (GL) 95-05 NDE and growth allowances for an example, the NDE

uncertainty component of 20% and a voltage growth allowance of 30%

per full power year can be utilized to establish a

VURL of 5.7 volts. The 20% NDE uncertainty represents a

square-root-sum-of-the-squares (SRSS) combination of probe wear

uncertainty and analyst variability. The degradation growth

allowance should be an average growth rate or 30% per effective full

power year, whichever is larger.

Relative to the expected leakage during accident condition

loadings, it has been previously established that a postulated MSLB

outside of containment but upstream of the main steam isolation

valve (MSIV) represents the most limiting radiological condition

relative to the plugging criteria. In support of implementation of

the revised plugging limit, analyses will be performed to determine

whether the distribution of cracking indications at the tube support

plate intersections during future cycles are projected to be such

that primary-to-secondary leakage would result in postulated site

boundary and control room doses exceeding 10 CFR 100, 10 CFR 50

Appendix A, and GDC-19 [General Design Criterion-19] requirements,

respectively. A separate calculation has determined the maximum

allowable MSLB leakage limit in a faulted loop. This limit was

calculated using the technical specification reactor coolant system

(RCS) Iodine-131 activity level of 1.0 microcuries per gram dose

equivalent Iodine-131 and the recommended Iodine-131 transient

spiking values consistent with NUREG-0800. The projected MSLB

leakage rate calculation methodology prescribed in Section 2.b of GL

95-05 will be used to calculate the end-of-cycle (EOC) leakage.

Projected EOC voltage distribution will be developed using the most

recent EOC eddy current results and considering an appropriate

voltage measurement uncertainty. The log-logistic probability of

[[Page 64110]]

leakage correlation will be used to establish the MSLB leakrate used

for comparison with the faulted loop allowable limit. Therefore, as

implementation of the voltage-based repair criteria does not

adversely affect steam generator tube integrity and implementation

will be shown to result in acceptable dose consequences, the

proposed amendment does not result in any increase in the

probability or consequences of an accident previously evaluated in

the Updated Final Safety Analysis Report (UFSAR).

The proposed changes to the BVPS-1 Index, Specifications and

associated Bases and the proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. Therefore, these changes do not involve an

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

Implementation of the proposed steam generator tube voltage-

based repair criteria does not introduce any significant changes to

the plant design basis. Use of the voltage-based repair criteria

does not provide a mechanism which could result in an accident

outside of the region of the tube support plate elevations as no

outside diameter stress corrosion cracking (ODSCC) is occurring

outside the thickness of the tube support plates. Neither a single

or multiple tube rupture event would be expected in a steam

generator in which the plugging limit has been applied (during all

plant conditions).

Duquesne Light Company will implement a maximum primary-to-

secondary leakage rate limit of 150 gpd [gallons per day] per steam

generator to help preclude the potential for excessive leakage

during all plant conditions. The RG 1.121 criterion for establishing

operational leakage rate limits that require plant shutdown are

based upon leak-before-break considerations to detect a free span

crack before potential tube rupture during faulted plant conditions.

The 150 gpd limit provides for leakage detection and plant shutdown

in the event of the occurrence of an unexpected single crack

resulting in leakage that is associated with the longest permissible

crack length. RG 1.121 acceptance criteria for establishing

operating leakage limits are based on leak-before-break

considerations such that plant shutdown is initiated if the leakage

associated with the longest permissible crack is exceeded.

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

at 1.43 times the MSLB pressure differential and the MSLB pressure

differential alone are approximately 0.57 inch and approximately

0.84 inch, respectively. A leak rate of 150 gpd will provide for

detection of approximately 0.41 inch long cracks at nominal leak

rates and approximately 0.62 inch long cracks at the lower 95%

confidence level leak rates. Since tube burst is precluded during

normal operation due to the proximity of the TSP to the tube and the

potential exists for the crevice to become uncovered during MSLB

conditions, the leakage from the maximum permissible crack must

preclude tube burst at MSLB conditions. Thus, the 150 gpd limit

provides for plant shutdown prior to reaching critical crack lengths

for MSLB conditions using the lower 95% leakrate data. Additionally,

this leak-before-break evaluation assumes that the entire crevice

area is uncovered during blowdown. Partial uncovery will provide

benefit to the burst capacity of the intersection. Analyses have

shown that only a small percentage of the TSPs are deflected greater

than the TSP thickness during a postulated MSLB.

As steam generator tube integrity upon implementation of the

voltage-based repair criteria continues to be maintained through

inservice inspection and primary-to-secondary leakage monitoring,

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

accident previously evaluated is not created.

The proposed change to BVPS-1 Index, Specifications and

associated Bases and the proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. These changes do not change the performance of

plant systems, plant configuration or method of operating the plant.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The use of the voltage-based repair criteria at BVPS-2 maintains

steam generator tube integrity commensurate with the criteria of RG

1.121. This guide describes a method acceptable to the Commission

for meeting GDCs 14, 15, 30, 31, and 32 by reducing the probability

or the consequences of steam generator tube rupture. This is

accomplished by determining the limiting conditions of degradation

of steam generator tubing, as established by inservice inspection,

for which tubes with unacceptable cracking should be repaired or

removed from service. Upon implementation of the proposed criteria,

even under the worst case conditions, the occurrence of ODSCC at the

tube support plate elevations is not expected to lead to a steam

generator tube rupture event during normal or faulted plant

conditions. The EOC distribution of crack indications at the tube

support plate elevations will be confirmed to result in acceptable

primary-to-secondary leakage during all plant conditions and that

radiological consequences remain within the licensing basis.

In addressing the combined effects of loss-of-coolant-accident

(LOCA) + safe shutdown earthquake (SSE) on the steam generator

component (as required by GDC 2), it has been determined that tube

collapse may occur in the steam generators at some plants. This is

the case as the tube support plates may become deformed as a result

of lateral loads at the wedge supports at the periphery of the plate

due to the combined effects of the LOCA rarefaction wave and SSE

loadings. Then, the resulting pressure differential on the deformed

tubes may cause some of the tubes to collapse. There are two issues

associated with steam generator tube collapse. First, the collapse

of steam generator tubing reduces the RCS flow area through the

tubes. The reduction in flow area increases the resistance to flow

of steam from the core during a LOCA which, in turn, may potentially

increase peak clad temperature. Second, there is a potential that

partial through-wall cracks in tubes could progress to complete

through-wall cracks during tube deformation or collapse.

The results of an analysis using the larger break inputs show

that the LOCA loads were found to be of insufficient magnitude to

result in steam generator tube collapse or significant deformation.

Since the leak-before-break methodology is applicable to the reactor

coolant loop piping, the probability of breaks in the primary loop

piping is sufficiently low that they need not be considered in the

structural design of the plant. The limiting LOCA event becomes the

pressurizer spray line break. Analysis results have demonstrated

that no tubes were subject to deformation or collapse. No tubes have

been excluded from application of the subject voltage-based steam

generator tube repair criteria.

Addressing RG 1.83 considerations, implementation of the

voltage-based repair criteria is supplemented by: enhanced eddy

current inspection guidelines to provide consistency in voltage

normalization, the bobbin coil inspection will include 100% of the

hot-leg TSP intersections and cold-leg intersections down to the

lowest cold-leg TSP with known ODSCC, the determination of the TSPs

having ODSCC will be based on the performance of at least 20% random

sampling of tubes inspected over their full length, and rotating

pancake coil inspection requirements for the larger indications left

inservice to characterize the principal degradation as ODSCC.

As noted previously, implementation of the tube support plate

intersection voltage-based repair criteria will decrease the number

of tubes which must be repaired. The installation of steam generator

tube plugs reduces the RCS flow margin. Thus, implementation of the

voltage-based repair criteria will maintain the margin of flow that

would otherwise be reduced in the event of increased tube plugging.

The proposed change to the BVPS-1 Index, Specifications and

associated Bases and the proposed change to BVPS-2 Table 4.4-2 are

editorial in nature. These changes will not reduce the margin of

safety because they have no impact on any safety analysis

assumptions.

Based on the above, it is concluded that the proposed license

amendment request does not result in a significant reduction in

margin with respect to plant safety as defined in the UFSAR or any

BASES of the plant technical specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

Potts &

[[Page 64111]]

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

NRC Project Director: Robert A. Capra.

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 amendment request: October 15, 1998.

Description of amendment request: The proposed amendment would make

several changes that are administrative in nature. The changes would

(1) make editorial changes to delete obsolete material or material

adequately described elsewhere, change action statement numbers, update

the technical specification (TS) index pages, and make changes to be

consistent with the guidance of the improved standard technical

specifications (ISTS); (2) delete reporting requirements that duplicate

reporting requirements contained in 10 CFR; and (3) relocate the

requirement for meteorological monitoring instrumentation from the TS

to the Licensing Requirements Manual.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

a. This change deletes an expired Unit 1 license condition and a

Unit 2 license requirement that is not required since it is

redundant to the reporting requirements addressed in 10 CFR 50.73.

Deleting these requirements does not involve any increase in the

probability or consequences of an accident previously evaluated.

b. The reference to Specification 3.0.6 was omitted from

Specification 3.0.1 in Unit 1 Amendment 213 and Unit 2 Amendment 90

and is being added to 3.0.1 to be consistent with the Improved

Standard Technical Specifications of NUREG 1431. This does not

involve any increase in the probability or consequences of an

accident previously evaluated.

c. The Core Alteration definition has been updated to be

consistent with the regulations and ISTS. The Offsite Dose

Calculation Manual (ODCM) definition has been updated to be

consistent with the change to Administrative Control 6.9.3. The

Members of the Public definition has been changed to be consistent

with 10 CFR 20.1003. This does not involve any increase in the

probability or consequences of an accident previously evaluated.

d. Changing Table 3.3-6 Action Statement 36 to Action Statement

35 is an editorial change to eliminate redundant use of action

statement numbers. This does not involve any increase in the

probability or consequences of an accident previously evaluated.

e. The technical specification index is being revised to address

removal of the Meteorological Monitoring specification and title and

page number changes to the administrative control reporting

requirements section. The Meteorological Monitoring specification is

being relocated to the Licensing Requirements Manual (LRM).

Relocating the Meteorological Monitoring requirements is in

accordance with the guidance in the Commission's Final Policy

Statement and revisions to 10 CFR 50.36 on the content of the

technical specifications and the ISTS. The Meteorological Monitoring

requirements do not meet any of the criteria, 1 thru 4 of 10 CFR

50.36 and can, therefore, be relocated from the Technical

Specifications to the LRM. These changes do not involve any increase

in the probability or consequences of an accident previously

evaluated.

f. The exclusion area boundary is adequately described in each

unit's UFSAR [Updated Final Safety Analysis Report], therefore,

design feature 5.1 Site Location is also being modified by deleting

the description of the exclusion area boundary. This does not

involve any increase in the probability or consequences of an

accident previously evaluated.

g. The change to refer to the Unit 1 Overpressure Protection

System (OPPS) enable temperature in Specification 3.4.9.3 in lieu of

specifying 275 deg.F was evaluated and found acceptable in the

request for approval of Amendment 160. The deletion of the asterisk

in Unit 2 Specification 3.9.8.1 was justified as part of the request

for approval of Amendment 25. The inadvertent omission of the ACTION

to take in the case that the temperature of the steam generator is

precisely 50 deg.F above the cold leg temperature is being

corrected. The cases of greater than and less than 50 deg.F are

already included. These are editorial changes that do not involve

any increase in the probability or consequences of an accident

previously evaluated.

h. The administrative control reporting requirements have been

modified to incorporate various ISTS requirements. This requires

changing titles and eliminating requirements addressed elsewhere,

removing reference to deleted sections, and replacing reference to

the administrative control section reporting requirements in various

specifications with reference to 10 CFR 50.4. The 1993 NRC final

policy statement set forth the criteria for determination of those

requirements to be included in TS. The reporting requirements being

removed from the TS do not meet the criteria for inclusion in the

TS; therefore, the reporting requirements have been modified to

reflect those requirements provided in the ISTS. These are editorial

changes that do not involve any increase in the probability or

consequences of an accident previously evaluated.

i. The Technical Specification index has been modified to

address the revised pages.

These changes have been determined to be editorial and

administrative in nature, and as such, would not affect any accident

assumptions or radiological consequences of an accident. Therefore,

the proposed changes would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The editorial changes, the elimination of reporting requirements

which duplicate 10 CFR requirements and administrative improvements

to incorporate the ISTS requirements are all changes that are

administrative in nature. The proposed changes will not affect any

plant system or structure, nor will they affect any system

functional or operability requirements. Consequently, no new failure

modes are introduced as a result of the proposed changes. Therefore,

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

different type of accident from any accident previously evaluated.

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

of safety?

The proposed amendment modifies reporting requirements and

incorporates associated editorial changes that do not impact the

UFSAR design basis or accident analyses assumptions. This change

does not introduce any new operational modes or physical

modifications to the plant; therefore, no action will occur that

will involve a significant reduction in a margin of safety. In

addition, the proposed change does not affect radiological release

limits, monitoring equipment or operating practices. Therefore, the

proposed amendment does not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: Robert A. Capra.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: September 23, 1998.

Description of amendment request: The proposed amendment would

change Division III battery specific gravity acceptance criteria

outlined in River Bend Station (RBS) Technical Specifications (TS). The

change is required as a result of battery system design modifications

which are scheduled to be implemented in April 1999 during refueling

outage (RF) RF-8. During this time, the current Division III

[[Page 64112]]

battery will be replaced. The new battery, which also will have a

greater capacity rating, will be supplied with a nominal specific

gravity of 1.215 at 77 deg.F in contrast to the existing Division III

battery supplied with a nominal specific gravity of 1.210 at 77 deg.F.

Since TS Section 3.8.6, Table 3.8.6-1 values for specific gravity are

based on the manufacturer's nominal specific gravity, these values will

need to be updated.

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.

The system loads, voltage requirements, and inrush currents have

been calculated in accordance with IEEE Std. 485, ``IEEE Recommended

Practice for Sizing Large Lead Storage Batteries for Generating

Stations and Substations.'' To support these design requirements at

a capacity of 80%, a new battery must be installed. The nominal

specific gravity of the new battery, as provided by the manufacturer

of the battery, is 1.215 at 77 deg.F.

A review of USAR Chapter 15, including Appendix 15A, was

conducted to determine what accidents, if any, may be impacted by

the proposed change to the Division III battery specific gravity.

USAR Sections 15.2, ``Increase in Reactor Pressure;'' 15.3,

``Decrease in Reactor Coolant System Flow Rate;'' and Section 15.6,

``Decrease in Reactor Coolant Inventory'' discuss accidents that

involve the initiation of HPCS when reactor vessel level drops to

the initiation point. The function of the HPCS System is to mitigate

the consequences of an accident (i.e., to maintain reactor vessel

coolant inventory after small breaks which do not depressurize the

reactor vessel, or provide spray cooling heat transfer following

larger breaks, Ref. USAR Section 6.3.1.2.1). The function of the

Division III 125 Vdc Power System is to provide a highly reliable,

continuous, and independent source of control and motive power for

the HPCS System logic, HPCS diesel generator set control and

protection, and all Division III related control (Ref. USAR Section

8.3.2.2.1). This is a support function for the HPCS System.

USAR Section 15.5, ``Increase In Reactor Coolant Inventory,''

postulates an inadvertent HPCS actuation resulting from operator

error. The proposed changes to the Division III battery specific

gravity cannot result in an inadvertent HPCS actuation/injection.

The proposed changes to the allowable specific gravity values

provided in Technical Specification 3.8.6 are in agreement with the

manufacturer's nominal specific gravity. The revision simply ensures

that the battery has sufficient capacity to meet the energy

requirements of its critical loads. The proposed change does not

create any new internally generated missiles, nor does it affect the

High Energy Line Break Analysis or any other accident described in

Chapter 15 of the USAR. Neither the function nor the operation of

the Division III battery is impacted by the proposed change.

The replacement Division III battery will be supplied by the

manufacturer with a nominal specific gravity of 1.215 at 77 deg.F.

The battery manufacturer's rated performance is based on the

specific gravity of the battery being maintained near the nominal

specific gravity. Since the Division III design basis calculation

depends on the battery manufacturer's rated performance, battery

parameters upon which that performance is based must be monitored.

The current Technical Specification values for specific gravity are

based upon a nominal specific gravity of 1.210 at 77 deg.F. The

proposed values accurately reflect the manufacturer's nominal

specific gravity. Testing the Division III battery to the proposed

values provides assurance that the HPCS functions supported by the

125 Vdc System will not be adversely affected by the Division III

battery.

The proposed changes will not affect failure modes of existing

equipment. The proposed changes do not affect the ability of any

structures, systems or components to perform their safety functions.

Therefore, no undue risk to the health and safety of the public has

been created by the proposed changes, nor is there any change in the

radiological consequences at the site boundary.

By incorporating the correct value for battery specific gravity

verification in Table 3.8.6-1, the Technical Specifications will

accurately reflect the new design basis value for the Division III

battery specific gravity. This change allows the performance of the

Division III battery to be verified against the correct design basis

value, thus providing assurance that the Division III 125 Vdc power

system function will remain as assumed in the accident analysis.

Therefore, the proposed change cannot affect any accidents

previously evaluated (probability or consequences). Consequently,

the proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. This request does not create the possibility of occurrence of

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

Since a battery's capacity decreases as specific gravity

decreases below the manufacturer's nominal value, monitoring the

battery's specific gravity is one means of ensuring that the battery

will adequately supply the minimum energy required to support the

system function assumed in the accident analysis.

All safety systems will continue to function as originally

designed. The subject equipment will not function in a manner

different than described in USAR Section 8.3.2.2. The functional and

performance requirements of the Division III 125 Vdc System and its

associated interfaces have not been altered. The proposed change

simply ensures that the HPCS battery performance is verified against

the correct design basis value. This value provides assurance that

the HPCS System functions will not be adversely affected by the

capacity of the battery. Therefore, the proposed changes do not

create the possibility of occurrence of a new or different kind of

accident from any previously evaluated.

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

margin of safety.

This proposed change updates the acceptance criteria of the

current specific gravity for the Division III battery. This

acceptance criteria is in accordance with manufactures

recommendations. The design and license basis for the Division III

systems and functions remain unchanged and the battery will continue

to supply the 125 Vdc loads necessary to support these functions.

This value will reflect the manufacturer's nominal specific gravity

for the Division III battery. With the system functions supported as

assumed in the accident analyses, the margin to safety remains

unchanged.

As a result, the proposed change does not involve a significant

reduction in a margin to 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 are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

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

1400 L Street, NW, Washington, DC 20005.

NRC Project Director: John N. Hannon.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 8, 1998.

Description of amendment request: The proposed amendment would

implement Boiling Water Reactor Owners Group (BWROG) Enhanced Option I-

A (EIA) Reactor Stability Long Term Solution as documented in NEDO-

32339, Revision 1, ``Reactor Stability Long-Term Solution, Enhanced

Option I-A.'' The EIA long term solution has been accepted by the NRC

in Safety Evaluation Report, ``Reactor Stability Long-Term Solution,

Enhanced Option I-A Generic Technical Specifications (TS), NEDO-32339,

Supplement 4.''

The proposed changes to the RBS TS will enable the full

implementation of the Enhanced Option I-A (EIA) long term solution to

the neutronic/thermal hydraulic instability issue. Specifically, the

proposed change deletes the limits

[[Page 64113]]

on power and flow conditions associated with the implementation of the

guidance in General Electric Service Information Letter #380, Revision

1, ``BWR Core Thermal Hydraulic Stability'' (current TS 3.4.1, Figure

3.4.1-1 and RBS plant procedures), adds new specifications, to

establish limits for Fraction of Core Boiling Boundary (FCBB) and the

Period Based Detection System (PBDS), modifies the RPS instrumentation

specification and the description of the contents of the Core Operating

Limits Report (COLR) in current TS 5.6.5. The two new specifications

require maintaining stability control and the availability of a

stability detection system during operation in defined regions of the

power and flow operating domain.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendments do no involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendments allow the implementation of the Enhanced

Option I-A (EIA) long term solution to the neutronic/thermal

hydraulic instability issue. Current TS restrictions on power and

flow conditions, number of operating recirculation loops, and

operator actions implemented to reduce the probability of neutronic/

thermal hydraulic instability are eliminated and new stability

requirements consistent with NEDO-32339-A, Supplement 4, Revision 1,

are imposed.

While the proposed amendments permit operation in regions of the

power and flow operating domain postulated to be susceptible to

neutronic/thermal hydraulic instability, the implementation of the

EIA solution ensures there is not a significant increase in the

probability or consequences of an accident previously evaluated.

Operation in these regions does not increase the probability of

occurrence of initiators and precursors of other previously analyzed

accidents. The proposed amendments permit the implementation of the

features of the EIA solution which prevent neutronic/thermal

hydraulic instability. The features include pre-emptive reactor

scram upon entry into the regions of the power and flow operating

domain most susceptible to neutronic/thermal hydraulic instability--

the Exclusion Region. The EIA solution prevents neutronic/thermal

hydraulic instability during operation in regions of the power and

flow operating domain previously excluded from operation and

therefore does not significantly increase the probability of a

previously analyzed accident.

The EIA solution also requires implementation of stability

control prior to entry into a region of the power and flow operating

domain which is potentially susceptible, in the absence of stability

control, to neutronic/thermal hydraulic instability. The modified

rod block functions providing the restricted region entry alarm

(RREA), boiling boundary limits, and PBDS functions are required on

entry into the Restricted Region of the power to flow map. The

boiling boundary limits, and Period Based Detection System (PBDS)

functions are required on entry into the Monitored Region of the

power to flow map. The EIA solution prevents or allows for detection

and suppression of neutronic/thermal hydraulic instability during

operation in these regions of the power and flow operating domain.

The EIA solution includes restrictions on power and flow

conditions and actions associated with the modified APRM flow biased

scram and RREA functions. Required actions include adherence to the

boiling boundary limit stability control prior to entry and during

operation in the region of the power and flow operating domain which

is potentially susceptible to neutronic/thermal hydraulic

instability--in the absence of stability control. In addition, the

proposed amendments require operator actions based upon control room

indications generated by a new PBDS. The PBDS is designed to provide

alarm indication that conditions consistent with a significant

degradation in the stability performance of the reactor have

occurred and the potential for imminent onset of neutronic/thermal

hydraulic instability may exist. The PBDS also provides analog

indication of the highest and second highest successive period

confirmation count for all of the LPRMs monitored. This provides the

plant operators with continuous indication of reactor stability

operating conditions. The PBDS system provides indication only and

does not affect plant structures, systems, or components in any way

that could increase the probability or consequences of an accident.

Rather, the improved control room indications provide the operator

with more accurate and timely information.

The EIA solution allows for the ``Setup'' of APRM flow biased

scram and control rod block function. The EIA solution requires

adherence to certain boiling boundary limit stability controls prior

to selection by the operator of APRM flow biased scram and control

rod block function ``Setup'' setpoints. This ``Setup'' function

allows operation in a region of the power and flow operating domain

potentially susceptible to neutronic/thermal hydraulic instability

provided the additional limits of the flow control boiling boundary

(FCBB) and PBDS are met. After exiting the region requiring the

stability control to be met, the setpoints can be manually reset to

their normal values. Stability controls are required to be in place

when setpoints are ``Setup''. As a backup EIA feature, the APRM flow

biased setpoints automatically reset to their normal values above a

pre-determined flow condition. This automatic reset to the more

conservative setpoints ensures that the pre-emptive reactor scram

will prevent operation as a result of an anticipated operational

occurrence in the region most susceptible to neutronic/thermal

hydraulic instability should the operator not select the more

conservative setpoints appropriate for operation following exit from

the region requiring stability control. The FCBB, PBDS, and

automatic reset of the APRM flow biased scram and control rod block

function ``setup'' setpoints allow for the use of the ``setup''

feature and help ensure that there is not an increase in the

probability or consequences of an accident.

Operation in the regions of the power and flow operating domain

excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a

result of anticipated operational occurrences. In the absence of

operator actions the severity of these anticipated operational

occurrences may increase due to the potential occurrence of

neutronic/thermal hydraulic instability as a result of operation in

these regions. Upon entry, as a result of an anticipated operational

occurrence, into the region most susceptible to neutronic/thermal

hydraulic instability the pre-emptive reactor scram prevents

neutronic/thermal hydraulic instability. Therefore, the consequences

of an accident do not significantly increase while operating with

stability control in place.

The required EIA features is designed to limit possible

neutronic/thermal hydraulic instabilities and to detect and suppress

further neutronic/thermal hydraulic instabilities. These features

include: a pre-emptive automatic scram, the control rod block and

alarms associated with entry into the region susceptible to

neutronic/thermal hydraulic instabilities, automatic reset of APRM

flow biased setpoints, PBDS, FCBB, and the required operator

actions, including manual reactor scram. Therefore, the proposed

amendments prevent the occurrence of neutronic/thermal hydraulic

instability during operation or as a consequence of an anticipated

operational occurrence and do not significantly increase the

consequences of any previously analyzed accident.

2. The proposed amendments do not create the possibility of a

new or different kind of accident from any previously evaluated.

The proposed amendments eliminate existing restrictions on power

and flow conditions and impose alternative restrictions which permit

the implementation of the EIA long term stability solution. The

current restrictions on the power and flow conditions do not prevent

entry into regions of the power and flow operating domain most

susceptible to neutronic/thermal hydraulic instability and therefore

the possibility of neutronic/thermal hydraulic instability exists in

the absence of operator action. The required features of the EIA

solution implement a pre-emptive scram upon entry into the region

most susceptible to neutronic/thermal hydraulic instability, without

operator action. The accessible operating domain allowed by the

proposed amendments is essentially a subset of the power and flow

operating domain currently allowed. Initial conditions are bounded

by the current initiators and precursors of accidents and

anticipated operational occurrences. Accordingly, no new accident of

initiator is present. Therefore, the proposed amendments do not

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

that previously evaluated.

Concurrent with the implementation of the proposed amendments, a

modified Flow

[[Page 64114]]

Control Trip Reference (FCTR) card, EIA FCTR card, and a new Period

Based Detection System (PBDS) will be installed as required by the

EIA solution. The function of the EIA FCTR card is to aid the

operator in the identification of entry into regions of the power

and flow operating domain potentially susceptible to neutronic/

thermal hydraulic instability in the absence of stability controls

and to initiate a pre-emptive scram upon entry into the regions most

susceptible to neutronic/thermal hydraulic instability. This is

accomplished by altering the existing values of setpoints of the

APRM flow biased scram and the control rod block functions generated

by the EIA FCTR card.

The design of the EIA digital FCTR card is a functional

equivalent of the original analog FCTR card. The Failure Modes and

Effects Analysis (FMEA) for the card detailed in NEDC-32339P-A

Supplement 2 found no single failure that would increase the

consequences of an accident. The EIA FCTR card maintains the

original basis for the NMS interface functions of the analog FCTR

card it replaces. The plant specific environmental conditions

(temperature, humidity, pressure, seismic, and electromagnetic

compatibility) have been confirmed to be enveloped by the

environmental qualification values for the EIA FCTR cards.

Therefore, the potential for spurious scrams or common mode failures

induced by environmental effects (e.g., electromagnetic

interference) is considered negligible. The installation of the EIA

FCTR card will therefore not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The function of the PBDS is to provide the operator with an

indication that conditions consistent with a significant degradation

in the stability performance of the reactor has occurred and the

potential for imminent onset of neutronic/thermal hydraulic

instability may exist. This is accomplished by the installation of a

new PBDS card in the Neutron Monitoring System in accordance with

NRC approved BWROG and GE design. The PBDS card takes inputs from

individual local power range monitors and provides analog indication

of the highest and second highest successive period confirmation

count, provides a Hi DR and Hi-Hi DR alarm, and INOP status

indication to the operator in the control room. These displays can

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

from any accident previously evaluated. The plant specific

environmental conditions (temperature, humidity, pressure, seismic,

and electromagnetic compatibility) have been confirmed to be

enveloped by the PBDS environmental qualification values. Therefore,

the installation of the PBDS card will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed amendments do not involve a significant

reduction in the margin of safety.

The proposed amendments permit the implementation of the EIA

long term solution to the stability issue. Under certain conditions,

existing BWR designs are susceptible to neutronic/thermal hydraulic

instability. GDC 10 of 10 CFR 50, Appendix A, requires that

specified acceptable fuel design limits not be exceeded during

anticipated operational occurrences. General Design Criterion (GDC)

12 of 10 CFR 50, Appendix A, requires thermal hydraulic instability

to be prevented by design or be readily and reliably detected and

suppressed. When the design of the reactor system does not prevent

the occurrence of neutronic/thermal hydraulic instability,

instability is considered an anticipated operational occurrence. The

proposed amendments and the associated design modifications provide

automatic features and operational information to the Control Room

that replace the existing BWROG Interim Corrective Actions (ICAs).

Thus the EIA solution assures compliance with GDC-10 and GDC 12 by

providing for reliable detection and suppression and by the

prevention of neutronic/thermal hydraulic instability. This

therefore precludes neutronic/thermal hydraulic instability from

becoming a credible consequence of an anticipated operational

occurrence. As a result the margins of safety are maintained.

Analyses performed by the BWROG indicate that neutronic/thermal

hydraulic instability induced power oscillations could result in

conditions exceeding the MCPR SL prior to detection and suppression

by the current design of the Neutron Monitoring System and Reactor

Protection System. To ensure compliance with GDC 12 the BWROG

developed Interim Corrective Actions (ICAs) to enhance the

capability of the operator to readily and reliably detect and

suppress neutronic/thermal hydraulic instability. The BWROG ICAs

also provided additional guidance for monitoring local power range

monitors beyond the requirements of current TS 3.4.1 to ensure

adequate margin to the onset of neutronic/thermal hydraulic

instability. Reliance on operator actions to comply with GDC 12 was

accepted on an interim basis by the NRC pending final implementation

of a long term solution to the stability issue. The modified design

of the Reactor Protection System (APRM flow biased scram) and

stability control prior to entry into a region of the power and flow

operating domain which is potentially susceptible, in the absence of

stability control, to neutronic/thermal hydraulic instability

implemented with the EIA solution prevents neutronic/thermal

hydraulic instability. In addition, significant backup protection

features, including the PBDS and specified operator actions, are

required to be implemented. As a result, the margin to the onset of

neutronic/thermal hydraulic instability provided by the existing TS

requirements and BWROG ICAs recommendations is not reduced by the

implementation of the EIA solution. The EIA solution assures

compliance with GDC 12 by the prevention of neutronic/thermal

hydraulic instability and therefore precludes neutronic/thermal

hydraulic instability from becoming a credible consequence of an

anticipated operational occurrence. The consequences of anticipated

operational occurrences will not increase and the margin to the MCPR

SL will not decrease upon implementation of the EIA solution.

Therefore, the proposed amendment does not involve a 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: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

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

1400 L Street, NW, Washington, DC 20005.

NRC Project Director: John N. Hannon.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October 1, 1998.

Description of amendment request: The proposed change modifies

Technical Specification (TS) 3.3.3.7.3 and Surveillance Requirement

4.3.3.7.3 for the broad range gas detection system. A change to

Technical Specification Basis 3/4.3.3.7 has been included to support

this change. This change to the TS is necessary for the installation of

a new, more reliable broad range gas detection system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The broad range gas detection system has no effect on the

accidents analyzed in Chapter 15 of the Final Safety Analysis

Report. It's only effect is on habitability of the control room,

which will be enhanced by installation of the new monitoring system

and this change to the Technical Specifications. Qualitative

analysis based on a quantitative risk assessment has shown that the

impact on operator incapacitation and subsequent core damage risk of

the periodic automatic background/reference spectrum check is

negligible and that the probability of malfunction of the BRGMs due

to a slowly increasing toxic chemical concentration is negligible.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different type of

[[Page 64115]]

accident from any accident previously evaluated?

Response: No.

The proposed Technical Specification change in itself does not

change the design or configuration of the plant. The new broad range

toxic gas monitoring system performs the same function as the old

system, but it accomplishes this function with increased

reliability.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The broad range gas detection system has no effect on a margin

of safety as defined by Section 2 of the Technical Specifications.

Its only effect is on habitability of the control room, which will

be enhanced by installation of the new monitoring system and this

change to the Technical Specifications.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

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

L Street NW, Washington, DC 20005-3502.

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: July 30, 1998 (LAR-222).

Description of amendment request: The proposed amendment will

change the Improved Technical Specifications (ITS) to add a new

Required Action for the existence of breaches in the Control Complex

Habitability Envelope (CCHE) that are in excess of allowances. A new

surveillance requirement for the performance of a periodic integrated

leak test of the CCHE boundary on a 24-month frequency would also be

added. Changes to the current Ventilation Filter Test Program (VFTP)

are proposed to adopt current standards for laboratory testing, change

acceptable values of control room emergency ventilation flow rate and

filter differential pressure, and add the Auxiliary Building

Ventilation Exhaust Filters to the VFTP. Conforming changes to the ITS

Bases are also included.

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated. The Control Room Emergency Ventilation System (CREVS) and

the Control Complex Habitability Envelope (CCHE) are designed to

limit the radiation dose to the control room operating staff

following a design basis accident. Since these systems are only

effective in limiting dose following an accident, the existence of

limited breaches in the CCHE, the performance of periodic leak

tests, and changes to the Ventilation Filter Test Program (VFTP)

would not increase the probability of occurrence of any evaluated

event. The features of the CREVS and the Control Complex emergency

filters, or the CCHE have no direct function in mitigating the

offsite consequences of any evaluated accident. The Auxiliary

Building exhaust filters are not credited with reducing offsite

doses, however, if available would filter releases from the

Auxiliary Building. Adding them to the VFTP will not increase the

consequences calculated for any evaluated accident.

The proposed changes are consistent with the revised control

room operator dose calculations as presented in the Control Room

Habitability Report dated July 1998. Since all calculated doses are

within 10 CFR Part 50, Appendix A GDC 19 limits there is no

significant increase in consequences.

It is conceivable that the existence of additional breaches in

the CCHE could result in an increase in operator dose, however the

low probability of a catastrophic reactor accident, the relatively

short time allowed for breaches to be open in excess of approved

dose calculation assumptions, and the ability to close breaches

expeditiously makes the risk increase insignificant.

The changes to the ITS Bases improve information on the

operation and function of CREVS, and establish that CREVS

operability is dependent on maintaining CCHE integrity. The

inclusion of this information reinforces the importance of

maintaining the CCHE boundary, and will help to ensure the CREVS is

capable of performing its intended safety function.

The Control Room Habitability Report, dated July 1998, provided

with this LAR presents the methodology used and the results of the

operator dose calculations for the Maximum Hypothetical Accident,

toxic gas release, and other design basis accidents. The report

provides the information needed for NRC review of LAR 222, Revision

I and the associated unreviewed safety question. This evaluation

concludes that the current level of CCHE integrity provides adequate

protection for the control room operator.

Based on the foregoing, the proposed amendment does not

significantly increase the probability or consequence of an accident

previously evaluated.

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

accident from any accident previously evaluated.

Neither performance of periodic CCHE leak tests nor changes to

the existing VFTP can create the possibility of a new or different

kind of accident. During the period of time when CCHE breaches are

greater than the design calculation, there exists the possibility

that control room dose from an analyzed accident may be greater than

specified in General Design Criterion 19. This condition will not

however create the possibility of a new or different kind of

accident. Since CREVS and the emergency filtration units function to

provide protection following a radiological accident the changes

proposed to improve their performance cannot create a new or

different kind of accident. Changes to the Bases to provide better

information on determining CREVS and CCHE operability cannot create

the possibility of a new or different kind of accident.

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

safety.

The proposed amendment does not involve a significant reduction

in a margin of safety. Neither performance of periodic CCHE leak

tests nor changes to the existing VFTP can create a reduction in the

margin of safety. The changes to both of these programs will result

in improved assurance that the CREVS and CCHE will perform as

expected if required for operator protection. Changes to the Bases

of the CREVS Technical Specification which clarify the conditions

necessary for operability will improve understanding of the

requirements for maintaining control room habitability, and will not

create a reduction in the margin of safety. The existence of

additional breaches in the CCHE for short periods of time does not

significantly increase the risk of control room operator exposure to

airborne radioactivity or toxic gas. There is no change in the risk

to the public since the CCHE has no direct function in mitigating

the offsite consequences of any evaluated accident. Any event that

could create these exposures has an extremely low probability of

occurrence, and while the potential for higher operator exposure

exists if additional breaches are open, the short duration allowed

would not significantly increase the risk of exposure. Therefore,

for the reason stated above the existing margin of safety would not

be reduced.

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

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

satisfied.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

[[Page 64116]]

W. Crystal Street, Crystal River, Florida 34428.

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

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

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

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

Date of amendment request: September 30, 1998 (LAR-238).

Description of amendment request: The proposed amendment will

correct the reactor coolant system (RCS) leakage detection capability

of the Reactor Building atmosphere gaseous radioactivity monitor

described in the Improved Technical Specification Bases and the Final

Safety Analysis Report (FSAR). These documents currently identify that

the gaseous radioactivity monitor is capable of detecting a one gallon

per minute (gpm) RCS leak within one hour. The licensee has determined

that it would take approximately 14 hours for this instrument to detect

a one gpm RCS leak using currently accepted assumptions. The capability

of other monitors to detect a one gpm RCS leak within one hour is not

affected by this change.

The licensee cited several factors which contribute to the

difficulty in reliably detecting RCS leakage increases of one gpm

within one hour using a gaseous radioactivity monitor. These include

the relatively long half-life of Xe-133 (primary nuclide of detection),

fluctuations in background levels of radioactivity, the existence of

minor RCS leaks, improved performance of nuclear fuel, and improved

primary water chemistry control. Based on RCS radioactivity

concentrations assumed in the Environmental Report, half-lives of the

most abundant gaseous nuclides, and background radioactivity levels,

the licensee indicated a one gpm leak can conservatively be detected in

approximately 14 hours by the gaseous monitor. The licensee has

determined that this change to the licensing basis is an unreviewed

safety question.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No. The function of the RM-A6 gaseous radioactivity monitor is

to detect leakage from the RCS that may develop as a result of a

flaw in a pressure boundary component. The previously identified

capability to detect a one gpm leak within one hour would have

provided an earlier warning of a small RCS leak than the actual

detection capability now identified. However, RCS loss of coolant

accidents evaluated in the FSAR cover the full spectrum of break

sizes up to and including a complete severance of the largest RCS

piping. The results of these analyses demonstrate that the

consequences of such leaks are acceptable.

No other equipment relies on the capability of the RM-A6 gaseous

monitor's ability to detect RCS leakage to perform its function.

Likewise, no accident analyses rely on RCS leak detection for

successful mitigation. Identifying the detector's actual capability

to detect an RCS leak will not increase the probability of

occurrence of an RCS leak. Detection time for an RCS leak was a

consideration in granting a partial exemption to General Design

Criterion 4. However, the capability of the RCS piping to resist

propagation of a flaw from a leak into a break was based on material

fracture analysis and material properties, not on the ability to

detect low levels of leakage.

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

from any accident previously evaluated.

No. The function of the RM-A6 gaseous radioactivity monitor is

to detect RCS leakage that may develop from a flaw in a pressure

boundary component. The monitor is a passive component that provides

an indication of possible leakage for further operator evaluation.

Identifying that a longer response time is required for the monitor

to detect a small leak will not create the possibility of a new or

different kind of accident. Existing analyses for small and large

break loss of coolant accidents provide an evaluation of the full

spectrum of RCS break sizes.

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

No. The RM-A6 gaseous radioactivity monitor is included in plant

technical specifications as one of two containment atmosphere RCS

leak detection instruments required to be operable to satisfy a

limiting condition for operation. If the RM-A6 particulate monitor

is not operable, then the response time of the containment

atmosphere monitor will be increased. RCS piping analyses have

demonstrated that the propagation of a small primary loop leak into

a pipe break would not occur rapidly. NRC acceptance of the

applicable analyses included significant safety factors for the

propagation of flaws into pipe breaks which were based on low

probability stress combinations of normal plus safe shutdown

earthquake loads. Considering the actual detection capability of the

RM-A6 gaseous monitor and the existence of other diverse leak

detection capabilities, detection of a leak in a relatively short

period of time is anticipated. In the event an RCS leak developed

into a pipe break, current accident analyses would bound the effects

of the pipe break on and off site. Therefore, the possibility of

increased time to detect an RCS leak does not represent 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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

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

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

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

Date of amendment request: October 16, 1998 (LAR-229).

Description of amendment request: The proposed amendment would

change the Crystal River Unit 3 (CR-3) Final Safety Analysis Report

(FSAR), Improved Technical Specifications (ITS) and ITS Bases to

resolve an Unreviewed Safety Question (USQ). This USQ was created by

changing the normal standby position of valves DHV-34 and DHV-35 (low

pressure injection (LPI) pump suction valves from borated water storage

tank) from normally open to normally closed. Maintaining these valves

normally closed is necessary to ensure assumptions used in fire

protection analyses remain valid. The proposed amendment would also add

new ITS surveillance requirements for verifying on a periodic basis

that the LPI system components and piping, and the building spray

suction piping, are full of water.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

Valves DHV-34 and DHV-35 are located in the suction lines

between the borated water storage tank (BWST) and the low pressure

injection (LPI) and building spray (BS) pumps. These valves are

maintained normally closed, and are designed to automatically open

upon receipt of a reactor coolant system (RCS) low-low pressure

signal

[[Page 64117]]

of 500 psig or a reactor building (RB) high pressure signal of 4

psig from the engineered safeguards actuation system (ESAS). The

designed full stroke time of these valves is within the assumptions

of the accident analyses performed for the specific design basis

accidents that require the LPI and/or BS systems for accident

mitigation. This is the original design basis for these valves.

Therefore, the valves are fully capable of performing their intended

safety functions while being maintained normally closed.

The failure of one of these valves to open does not impact the

mitigation of previously analyzed accidents that require the

operation of the LPI and/or BS systems, and cannot increase the

probability of these accidents occurring. No RCS or secondary system

pressure boundaries are compromised, no release paths for

radioactive materials are created, and no challenge to any safety

limit or acceptance limit are created by maintaining these valves

normally closed.

A single, active failure causing one of these valves to fail to

open upon demand would render one train of LPI and BS unavailable

for accident mitigation. However, the accident analyses have already

accounted for the possibility of only one train of LPI and BS being

available, and the consequences of previously evaluated accidents

would therefore remain unchanged.

Undetected voiding in the LPI piping and components, and BS

suction piping, is highly unlikely to occur. Based on the design and

physical layout of the LPI system and BS system, and the monitoring

of the systems performed on a periodic basis, any potential for LPI

piping and components and BS suction piping voiding will be quickly

and easily recognized and corrected. Therefore, since voiding is not

likely to occur, the consequence of previously evaluated accidents

would not be significantly increased by the proposed change.

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

from previously evaluated accidents?

Failure of either valves DHV-34 or DHV-35 to open upon demand on

an ESAS signal will not create the possibility of a new or different

kind of accident. The LPI system and BS system are maintained in a

standby condition during normal plant operations, and automatically

actuate only after an accident has occurred to mitigate the effects

of the initiating accident. No RCS or secondary system pressure

boundaries are compromised, no release paths for radioactive

materials are created, and no challenges to any safety limit or

acceptance limit are created by maintaining these valves normally

closed. Additionally, the possibility of undetected voiding in the

LPI piping and components, and BS suction piping, is not likely to

occur by maintaining these valves normally closed. Therefore,

maintaining valves DHV-34 and DHV-35 normally closed will not be an

initiator of a new or different kind of accident from previously

evaluated accidents.

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

Maintaining valves DHV-34 and DHV-35 normally closed will not

create a reduction in the margin of safety. Maintaining valves DHV-

34 and DHV-35 normally closed will ensure the capability to safely

shut down the reactor under certain postulated fire scenarios, but

will result in an extremely small increase in the probability of

failure of one train of LPI and BS to perform its safety functions.

Based on use of the CR-3 Probabilistic Safety Analysis (PSA) model,

and assuming the failure of either valve DHV-34 or DHV-35 to open,

the impact on the core-damage frequency was estimated and determined

to slightly increase from 7.38 E-6 to 7.41 E-6 per year. This

increase (3 E-8 or 0.4%) is in the range considered acceptable in

Regulatory Guide 1.174, ``An Approach for Using Probabilistic Risk

Assessment in Risk-Informed Decisions on Plant-Specific Changes to

the Current Licensing Basis,'' dated July 1998.

Maintaining these valves normally closed will not result in

undetected voiding in the LPI piping and components, and BS suction

piping, as a result of performance of periodic pressure monitoring.

If voiding occurs, the Improved Technical Specifications specify the

actions required to restore the affected systems to operable status,

including correcting the external leakage creating the observed

pressure decay. Therefore, the proposed monitoring will ensure the

margin of safety is not reduced.

Based on these benefits and risks, there is no discernible

change in the risk to the public in mitigating the offsite

consequences of any evaluated accident since the failure of one

train of LPI and/or BS for any reason is bounded by the assumptions

of the accident analyses. Failure of valve DHV-34 or DHV-35 to open

upon demand results in extremely low increases in the potential for

reactor core damage. Therefore, the existing margin of safety will

not be reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

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

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: October 15, 1998.

Description of amendment request: The proposed amendment request

would revise the TMI-1 Updated Final Safety Analysis Report (UFSAR)

Chapter 14 postulated accident analysis radiological dose consequences

resulting from application of revised atmospheric dispersion factors

(X/Q) at the Technical Specification Section 5.1.1 defined exclusion

area boundry (EAB) and low population zone (LPZ).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed amendment has no effect on

structures, systems or components. More extensive and recent

meteorological data have been utilized for atmospheric dispersion

factor (X/Q) determination for both EAB and LPZ. An evaluation of

the design basis accidents with revised EAB and LPZ X/Q values

results in increases in UFSAR Chapter 14 EAB and LPZ dose

consequences which remain well within the guidelines of 10 CFR Part

100.

Therefore, this activity does not involve a significant increase

in the probability of occurrence or the consequences of an accident

previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. The proposed

amendment has no impact on any plant structures, systems or

components. The proposed change revises the atmospheric dispersion

factors for EAB and LPZ used in the existing UFSAR Chapter 14

accident analyses, based on more extensive meteorological data.

These changes only effect the postulated dose consequences of

currently analyzed accidents. Therefore, this activity does not

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

any previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment has no impact on structures, systems

or components. The proposed revisions to the EAB and LPZ X/Q values

are based on recent more extensive meteorological data and

Regulatory Guide 1. 145 methods. The increased X/Q values provide a

more accurate assessment of meteorological conditions which result

in postulated dose consequences which remain well within the

guidelines of 10 CFR Part 100. Therefore, this activity does not

reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

[[Page 64118]]

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: Cecil O. Thomas.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: October 19, 1998.

Description of amendment request: The proposed Technical

Specification change request would add operability and surveillance

requirements for the remote shutdown system similar to those in NUREG-

1430, ``Standard Technical Specifications--Babcock and Wilcox Plants''

Section 3.3.18 entitled ``Remote Shutdown System''.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed amendment adds operability and

surveillance requirements for the existing TMI-1 remote shutdown

system similar to those contained in NRC NUREG-1430, ``Standard

Technical Specifications--Babcock & Wilcox Plants''. The addition of

these requirements to Technical Specifications provides further

assurance of remote shutdown system operability in the event that

operators must place and maintain the unit in a safe shutdown

condition from outside the control room. The function and operation

of the remote shutdown system has not changed. Therefore, this

activity has no affect on the probability of occurrence or

consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. The proposed

amendment has no impact on any plant structures, systems or

components. The function and operation of the remote shutdown system

has not changed. Therefore, this activity does not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment provides additional assurance of

remote shutdown system operability. The function and operation of

the remote shutdown system has not changed. Therefore, this activity

does not reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Cecil O. Thomas.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: October 19, 1998.

Description of amendment request: The proposed change to the TMI-1

Technical Specification would revise the limit on reactor coolant

system activity to a maximum allowable of 1.0 microcurie/gram dose

equivalent I-131. The proposed revision provides an allowable reactor

coolant system specific activity limit base on once-through steam

generator (OTSG) inspection results performed each refueling outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed amendment has no effect on

structures, systems or components. The existing steam line break

criteria are maintained. This change only accounts for radiological

consequences resulting from a revised maximum allowable reactor

coolant system (RCS) specific activity limit of 1.0 iCi/gm.

The new radiological consequences of the revised MSLB accident,

which also incorporate more conservative values for atmospheric

dispersion, are below 10 CFR 100 limits and 10 CFR 50, Appendix A,

GDC-19 limits for the control room. The use of revised atmospheric

dispersion factors for other TMI-1 accident analysis is addressed in

a separate license amendment request submittal. Therefore, this

activity does not involve a significant increase in the probability

of occurrence or the consequences of an accident previously

evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. The proposed

amendment has no impact on any plant structures, systems or

components. OTSG tube structural integrity is maintained. Therefore,

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

kind of accident from any previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment has no impact on structures, systems

or components. OTSG tube structural integrity is maintained. The

existing TMI-1 Technical Specification Section 3.1.4.1 Bases state

that the limitations on the specific activity of the primary coolant

ensure that the resulting 2-hour doses at the site boundary will be

well within the Part 100 limit following associated design basis

accidents postulated in conjunction with an assumed steady state

primary-to-secondary steam generator tube leakage of 1.0 gpm. This

margin of safety is preserved since resulting does consequences

incorporating more conservative values for atmospheric dispersion

remain well within the Part 100 limit. Therefore, this activity does

not reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: Cecil O. Thomas.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of amendment request: October 23, 1998.

Description of amendment request: The proposed amendment would

allow implementation of a feedwater leakage control system to address

leakage through the primary containment feedwater penetration isolation

valves.

[[Page 64119]]

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change implements a method of providing a

qualified sealing system for the primary containment feedwater

penetration isolation valves. This water sealing function, i.e., the

FWLCS, constitutes a new operating mode of the Residual Heat Removal

(RHR) system. The FWLCS introduces new piping that constitutes an

extension of the reactor coolant system (RCS); however, such piping

is designed to the same requirements as other RCS piping and as such

introduces no significant increase in the probability of any

accident previously evaluated. Notwithstanding, a postulated line

break in any of the new FWLCS piping would not, by itself, introduce

any new effects or consequences not already bounded by postulated

line-break or LOCA events previously evaluated in the USAR. Since

the proposed change does not affect any parameters or conditions

that contribute to the initiation of any accidents previously

evaluated, the proposed change cannot increase the probability of

any accident previously evaluated.

The proposed change potentially affects the leak-tight integrity

of the primary containment designed to mitigate the consequences of

a loss-of-coolant accident (LOCA). Once the FWLCS mode has been

initiated and a water seal for the seating surfaces of the primary

containment feedwater penetration isolation valves has been

established (within one hour after the accident), post-LOCA primary

containment atmosphere will be prohibited from leaking through the

feedwater penetrations and thus bypassing the secondary containment.

Calculations of post-accident DBA LOCA doses affected by this

change use accepted ICRP 30 dose conversion factors and take credit

for suppression pool scrubbing. Suppression pool scrubbing is

effective in reducing iodine release but has no assumed effect on

the removal of noble gases. Since the methodology and assumptions

for scrubbing are acceptable to the NRC per the guidance in SRP

Section 6.5.5 and the values for decontamination factors are

conservative, considerable margin is preserved within the analysis.

However, these calculations show increases in some of the previously

evaluated post-accident doses when compared with dose calculations

performed as part of the initial plant licensing basis. Although

some of the newly calculated post-accident doses are larger than

those that were previously approved, the increases remain small

enough to be within the acceptance limits given in 10 CFR 50,

Appendix A, GDC 19 and in 10 CFR 100.11.

Since all of the newly calculated post-accident doses resulting

from the proposed addition of a water sealing system for the

feedwater primary containment penetration isolation valves are below

the 10 CFR 50, Appendix A, GDC 19 and 10 CFR 100.11 acceptance

limits, IP has concluded that the proposed change does not result in

a significant increase in the consequences of an accident previously

evaluated.

2. The proposed change institutes a new operating mode of the

RHR system (the FWLCS mode). When this mode is established, it will

reduce primary containment atmosphere leakage to the environment in

the event of a LOCA. Flow diverted from the RHR system to the FWLCS

has been evaluated, and has been determined to have no adverse

impact on the capability of the RHR system to perform its intended

safety functions. Further, the additional piping added for the FWLCS

is designed to appropriate requirements for the RCS, thus ensuring

that RCS integrity is maintained per design. Sufficient isolation

between the RCS and the RHR low-pressure piping will also be

maintained per the FWLCS design. Thus, no safety functions are

altered or impacted as a result of this change. Installing,

operating, or testing the components that support the FWLCS mode has

no influence on, nor does it contribute to the possibility of a new

or different kind of accident or malfunction from those previously

analyzed. Because the USAR analysis already assumes leakage through

the feedwater primary containment penetrations following a design

basis LOCA, and the subject change does not affect the type of

accident(s) that are postulated to occur, the proposed change does

not present the possibility of an accident of a different type.

Additionally, the change in dose analysis methodology does not

create an accident or malfunction of a different type since it only

involves the analysis of the effects of accidents or malfunctions

previously evaluated in the USAR.

Based on the above, IP has concluded that the proposed change

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

accident not previously evaluated.

3. The margin of safety impacted by the proposed change involves

the dose consequences of postulated accidents which are directly

related to the primary containment leakage rate, specifically those

consequences associated with dose attributable to leakage through

the feedwater lines which are secondary containment bypass leakage

paths.

Although considerable conservatisms were included in the

reanalysis, this reanalysis identified some dose values that

increased above the previously licensed values as well as some dose

values that decreased below the previously licensed values. However,

all of the radiation dose consequences resulting from the proposed

change will continue to be below the 10 CFR 50, Appendix A, GDC 19

and 10 CFR 100.11 acceptance criteria.

Except for providing a method of sealing the feedwater primary

containment penetration isolation valves (and therefore the method

of performing periodic leakage testing of these components) no other

change in the method of primary containment leakage testing or

secondary containment bypass leakage path testing is being proposed.

All other primary and secondary containment bypass leakage testing

will continue to be performed in accordance with existing Technical

Specification requirements. Adequate programs are in place to ensure

that proper maintenance and repairs are performed during the service

life of the primary containment, systems and components penetrating

the primary containment, and for all secondary containment bypass

leakage paths.

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

result in a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

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

Decatur, IL 62525.

NRC Project Director: Stuart A. Richards.

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

Unit No. 1, Washington County, Nebraska

Date of amendment request: September 28, 1998.

Description of amendment request: The proposed amendment request

would resolve an unreviewed safety question (USQ) and amend the

operating license to allow manual override capability for the

containment isolation actuation signal to reactor coolant system

letdown isolation 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, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed modification does not change the probability of any

accident previously evaluated since it does not change any mode of

normal operation. Neither the accident signal (CIAS) nor the

override feature is an initiator of an analyzed event. The

consequences of an accident are also not changed significantly due

to the fact that design and administrative controls ensure that

previous accident analyses are bounding. The associated isolation

valves will operate as they have in the past in response to an

accident signal. There is no single failure that would prevent the

letdown isolation function from occurring. The CIAS override feature

can only be used if operators have verified that an UHE is the event

which has taken place and safety functions are being met.

[[Page 64120]]

This ensures that no significant fuel failures will occur due to the

event and the consequences of overriding CIAS will not adversely

impact radiological conditions in the auxiliary building.

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

or different kind of accident from any accident previously

evaluated.

The proposed modification does not create any failure mode which

could impact the operation of the RCS or associated systems in a

manner that would create a new or different kind of accident. With

respect to the letdown isolation function, the plant will operate as

it previously has and will respond the same way, automatically, to

an accident signal. No new accidents have been identified.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The procedural restrictions associated with the use of the CIAS

override feature will ensure that existing analyses addressing the

consequences of an UHE will be bounding and that safety functions

will be maintained as defined in EOPs. The radiological consequences

of letdown restoration in the auxiliary building will be similar to

normal operating conditions and will be bounded by that assumed in

the EEQ analysis. RCS inventory and pressure control will be

maintained within the established procedural limits.

Letdown restoration capability already exists after ESF reset.

The modification permits letdown restoration to occur earlier than

it would previously have been possible.

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

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

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

amendment request involves no significant hazards consideration.

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

South 15th Street, Omaha, Nebraska 68102.

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

Street, NW, Washington, DC 20005-3502.

NRC Project Director: William H. Bateman.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: October 15, 1998.

Description of amendment request: The proposed Technical

Specification (TS) changes involve revising TS Section 3/4.10 to

include a new Special Test Exception allowing the reactor to be

considered in operational condition (OPCON) 4 (cold shutdown) during

inservice leak or hydrostatic testing with a reactor coolant water

temperature greater than 200 deg.F and less than or equal to 212 deg.F.

This is an exception to certain OPCON 3 (hot shutdown) requirements,

including primary containment. The proposed TS changes will permit

unrestricted access to the primary containment for the performance of

required inspections.

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 Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes do not make any physical alterations or

modifications to plant systems or equipment. The proposed TS changes

will permit the performance of inservice leak or hydrostatic

testing, with the reactor in OPERATIONAL CONDITION (OPCON) 4 (COLD

SHUTDOWN) and the average reactor coolant temperature greater than

200 deg.F and less than or equal to 212 deg.F. The probability of a

leak in the reactor coolant pressure boundary during inservice leak

or hydrostatic testing is not increased by considering the reactor

in OPCON 4 with reactor coolant temperatures greater than 200 deg.F

and less than or equal to 212 deg.F. The inservice leak and

hydrostatic testing is performed water solid or near water solid.

The stored energy in the reactor core will be very low and the

potential for failed fuel and a subsequent increase in reactor

coolant activity above TS limits is minimal. In addition, Secondary

Containment will be operable and capable of handling airborne

radioactivity from leaks that could occur during the performance of

inservice leak or hydrostatic testing. Requiring the Secondary

Containment to be operable will ensure that potential airborne

radioactivity from leaks will be filtered through the Standby Gas

Treatment System (SGTS), thereby limiting any radioactivity releases

to the environment.

In the event of a large primary system leak, the reactor vessel

would rapidly depressurize allowing the low pressure Emergency Core

Cooling System (ECCS) subsystems to operate. The capability of the

systems that are required for OPCON 4 would be adequate to keep the

core flooded under this condition. Small system leaks would be

detected by leakage inspections before significant inventory loss

has occurred. This is an integral part of the hydrostatic testing

program.

Therefore, the proposed TS changes will not significantly

increase the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes do not make any physical alterations or

modifications to plant systems or equipment. The proposed TS changes

do not adversely impact the operation of any plant equipment.

Allowing the reactor to be considered in OPCON 4 during hydrostatic

or inservice leak testing, with a reactor coolant temperature

greater than 200 deg.F and less than or equal to 212 deg.F, is an

exception to certain OPCON 3 (HOT SHUTDOWN) requirements, including

primary containment integrity. The hydrostatic or inservice testing

is performed water solid, or near water solid. The stored energy in

the reactor core will be very low and the potential for failed fuel

and a subsequent increase in coolant activity above TS limits is

minimal. In addition, the Secondary Containment will be operable and

capable of handling airborne radioactivity from leaks that could

occur during the performance of hydrostatic or inservice leakage

testing.

The inservice leak or hydrostatic test conditions remain

unchanged. The potential for a system leak remains unchanged since

the reactor coolant system is designed for temperatures exceeding

500 deg.F with similar pressures. There are no alterations of any

plant systems or components that cope with the spectrum of

accidents.

Therefore, the proposed TS changes will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes do not make any physical alterations or

modifications to plant systems or equipment. The proposed changes

will permit the performance of inservice leak and hydrostatic

testing with a reactor coolant temperature greater than 200 deg.F

and less than or equal to 212 deg.F and the reactor in OPCON 4.

Since the reactor vessel head will be in place, Secondary

Containment integrity will be maintained, and all systems required

in OPCON 4 will be operable in accordance with the applicable TS

requirements. The proposed TS changes will not have any significant

impact on any design basis accident or safety limit. The hydrostatic

or inservice leak testing is performed water solid, or near water

solid. The stored energy in the reactor core is very low and the

potential for failed fuel and a subsequent increase in coolant

activity would be minimal. In the event of a large primary system

leak, the reactor pressure vessel would rapidly depressurize and the

low pressure ECCS subsystems would function as designed to maintain

adequate reactor core coverage. This would ensure that the fuel

would not exceed peak clad temperature limits.

Also, requiring Secondary Containment integrity will assure that

potential airborne radioactive material can be filtered through the

SGTS. This will assure that any offsite doses remain well within the

limits of 10 CFR 100 guidelines. Small system leaks would be

detected by inspections before significant inventory loss could

occur.

Therefore, this proposed TS change will not involve a

significant reduction in a margin of safety.

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

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

[[Page 64121]]

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101.

NRC Project Director: Robert A. Capra.

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

Generating Station, Salem County, New Jersey

Date of amendment request: October 19, 1998.

Description of amendment request: The proposed amendment would

eliminate restrictions imposed by Technical Specification (TS) 3.0.4

for the Filtration, Recirculation and Ventilation System (FRVS) during

fuel movement and core alteration activities. Specifically, TS Limiting

Conditions for Operation (LCOs) 3.6.5.3.1 and 3.6.5.3.2 would each be

revised to add a note stating that the provisions of TS 3.0.4 are not

applicable for initiation of handling of irradiated fuel in the

secondary containment and core alterations provided that the plant is

in Operational Condition 5, with reactor water level equal to or

greater than 22 feet 2 inches.

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 TS change does not involve any physical changes to

plant structures, systems or components (SSC). FRVS will continue to

function as designed. FRVS is an Engineered Safety Feature (ESF)

designed to mitigate the consequences of an accident, and therefore,

can not contribute to the initiation of any accident. For refueling

accidents, the current design basis analysis of FRVS credits only

the iodine removal capability of the FRVS ventilation unit and

neglects the considerable iodine removal capability of the FRVS

recirculation units. In addition, this proposed TS change will not

increase the probability of occurrence of a malfunction of any plant

equipment important to safety, since the time limits imposed by the

current FRVS LCO Action Statements are not affected by these

proposed changes. The proposed changes merely allow entry into the

FRVS LCO Action Statement in order to support refueling activities.

Therefore, the proposed TS changes, which would permit the

initiation of core alterations and handling of irradiated fuel with

only one operable FRVS ventilation unit and four operable FRVS

recirculation units for a limited seven day period under specific

refueling conditions, would not result in the increase of the

consequences of an accident previously evaluated.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed TS changes do not involve any physical changes to

plant SSC. The design and operation of the FRVS is not changed from

that currently described in the [Updated Final Safety Analysis

Report] UFSAR. FRVS will continue to function as designed to

mitigate the consequences of an accident. No changes of any kind are

being made to FRVS, or its support or supported systems. Deleting

the restrictions imposed by TS 3.0.4 as proposed in this TS change

request eliminates a compliance restriction imposed by the current

TS. Since the current TS already provide a seven day period to

perform refueling activities with inoperable FRVS ventilation and

recirculation units, the proposed changes would not introduce plant

operation in a configuration that is not already permitted in the

TS. Therefore, there is no possibility that implementing this

proposed TS change would create a different type of malfunction to

the FRVS than any previously evaluated. In addition, the proposed TS

changes do not alter the conclusions described in the UFSAR

regarding operation of FRVS.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

in a margin of safety.

The proposed TS change involves the elimination of TS 3.0.4

restrictions imposed on the FRVS LCO. The TS 3.0.4 requirements

impose an unnecessary challenge to performing refueling activities

when the FRVS LCO Action Statements already sufficiently define the

remedial measures to be taken. The time limits imposed by the

current FRVS LCO Action Statements are not affected by these

proposed changes. The FRVS LCO will retain sufficient configuration

controls to appropriately maintain the capability of FRVS to

mitigate design basis refueling accidents, no new FRVS

configurations will be permitted by the proposed changes, and there

will be no reduction in any margin of safety resulting from this

proposed TS change. Therefore, the proposed TS change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Project Director: Robert A. Capra.

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: September 18, 1998.

Description of amendment request: The proposed amendment would

revise the VCSNS Technical Specifications (TS) to address the Best

Estimate Analyzer for Core Operations--Nuclear (BEACON) core power

distribution monitoring and support system. The BEACON system provides

continuous core monitoring capabilities to augment the flux mapping

system when rated thermal power (RTP) is greater than 25%. The proposed

amendment would also make editorial changes to TS 3.3.3.2 and 4.3.3.2.c

to delete the reference to Fxy.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change allows the Power Distribution Monitoring

System (PDMS) to be used for measuring power distribution limits

when Thermal Power is greater than 25% RTP. This includes relocating

manufacturing and measurement uncertainty values from the Technical

Specification to the COLR [core operating limit report]. Also

included in this change is the addition of a new specification and

bases section for the Power Distribution Monitoring System (PDMS).

The Technical Specification Power Distribution Limits are not being

changed; only the method in which they are measured is being

changed. The probability of an accident is not significantly

increased. The measurement of power distribution limits and the

location of manufacturing and measurement uncertainty values are not

initiators of any analyzed event. The change will not affect the

consequences of any analyzed event. The power distribution limits

will still be measured and verified to be within limits as required

by the current Technical Specification Surveillance. The cycle-

specific core operating limits, although not in Technical

Specifications, will be followed in the operation of VCSNS. The

actions as required by current Technical Specifications, when or if

limits are exceeded are not being

[[Page 64122]]

changed. This change will not significantly affect the assumptions

relative to the mitigation of accidents.

Each accident analysis addressed in the VCSNS Final Safety

Analysis Report will be examined with respect to changes in cycle-

dependent parameters, which are obtained from application of the

NRC-approved reload design methodologies, to ensure that the

transient evaluation of new reloads are bounded by previously

accepted analyses. This examination, which will be performed per

requirements of 10 CFR 50.59, ensures that future reloads will not

involve an increase in the probability or consequences of an

accident previously evaluated.

Therefore, the change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change allows the Power Distribution Monitoring

System (PDMS) to be used for measuring power distribution limits

when Thermal Power is greater than 25% RTP. This includes relocating

manufacturing and measurement uncertainty values from the Technical

Specification to the COLR. Also included is the addition of a new

specification and bases section for the Power Distribution

Monitoring System. No safety-related equipment, safety function, or

plant operation will be altered as a result of this proposed change.

No hardware is being added to the plant as part of the change. The

cycle specific variables are calculated using the NRC-approved

methods and submitted to the NRC to allow the Staff to continue to

trend the values of these limits. The Technical Specifications will

continue to require operation within the required core operating

limits and appropriate actions will be taken when or if limits are

exceeded. The change will not introduce any new accident initiators.

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

different kind of accident from any accident previously evaluated.

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

safety?

The proposed change allows the Power Distribution Monitoring

System (PDMS) to be used for measuring power distribution limits

when Thermal Power is greater than 25% RTP. The margin of safety

presently provided by current Technical Specifications remains

unchanged. Only the method in which the power distribution

measurements are obtained is being changed. This method is verified

by Westinghouse, and reviewed and approved by the NRC. Appropriate

measures exist to control the values of the manufacturing and

measurement uncertainties. The proposed amendment continues to

require operation within the core limits, as obtained from NRC-

approved reload design methodologies. Appropriate actions required

to be taken when or if limits are violated remain unchanged.

Future changes to measurement and manufacturing uncertainties

located in the current Technical Specification will be evaluated per

the requirements of 10 CFR 50.59. Since the 10 CFR 50.59 process

does not allow any reduction in the margin of safety, prior NRC

approval is required prior to a reduction in the margin of safety.

If the evaluation of the changes [does] not result in [an]

unreviewed safety question, prior NRC approval will not be required.

Additionally, the VCSNS Technical Specifications require that all

revisions of the plant COLR be submitted to the NRC upon issuance.

Therefore, the change does not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Project Director: Herbert N. Berkow.

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

364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston

County, Alabama

Date of amendment request: October 12, 1998.

Description of amendment request: The proposed amendments would

revise Section 6, ``Administrative Controls,'' of the current Units 1

and 2 Technical Specifications (TS) to recognize additional management

positions associated with the steam generator replacement project and

providing them the ability to approve procedures regarding this

project.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the FSAR [Final Safety Analysis Report]. The proposed changes have

no impact on the probability of an accident. The change being

proposed is administrative in nature and involves no physical

alteration of the plant or changes to setpoints or operating

parameters. The change will provide an appropriate level of review

and approval of procedures related to the FNP steam generator

replacement without impacting the operational attention of the

current on-site plant management. There is no change in the FNP

design basis as a result of this change and, as a result, does not

involve a significant increase in the consequences of an accident

previously evaluated.

(2) The proposed changes to the TS do not increase the

possibility of a new or different kind of accident than any 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.

(3) The proposed changes do not involve a significant reduction

in the margin of safety. Adding individuals with the appropriate

knowledge base to the list of individuals who can approve

procedures, which may affect plant nuclear safety, is administrative

in nature. There is no impact on the accident analyses. The training

and experience requirements for the newly designated management

positions are similar to those requirements for other FNP management

positions. Therefore the established level of procedure review and

approval is not adversely impacted. In addition, these changes allow

FNP management to remain focused on plant operations. Thus the

proposed changes do not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Project Director: Herbert N. Berkow.

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

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

Date of amendment request: September 28, 1998.

Description of amendment request: The proposed amendment would

modify the requirements applicable when one or more trains of fuel

handling building exhaust air or control room makeup and cleanup

filtration are inoperable, and eliminate the need to enter Technical

Specification 3.0.3 when multiple trains of these systems are

inoperable. In addition, the proposed changes would align the actuating

instrumentation and logic system required actions with those that are

applicable to the systems. Finally,

[[Page 64123]]

an administrative change is proposed to remove a footnote that is no

longer applicable to the facility.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes consist of:

(a) Assuring that the Specifications define consistent allowed

outage times when the same safety function is addressed in multiple

Specifications,

(b) Allowing a system to remain inoperable when appropriately

restrictive administrative controls are placed on operations that

could result in a challenge to the safety function of the system,

(c) Providing an appropriately short Allowed Outage Time for

inoperability needed to permit required maintenance and testing that

affects all trains of a system,

(d) Redefining system operability and associated actions in a

manner consistent with the system design and function,

(e) Aligning a system to the actuated condition on the loss of

an actuation channel,

(f) Using consistent terminology throughout the Specifications.

The proposed changes do not represent significant increases in

the probability or consequences of an accident because:

(a) The alignment of the action times between actuating system

and actuated system operability requirements do not affect the

probability or consequences since inoperability of the actuated

system has the same effect as inoperability of the actuating system.

Since the changes proposed to the actuating system action times will

reflect those of the actuated system action times, no change to the

allowed outage time applicable to the safety function addressed and

fulfilled by both, will occur.

(b) Administrative controls to prevent the conduct of operations

that could lead to a challenge to the safety function of the system

when the actuation system is inoperable, assures that the design

bases functions of the system will not be challenged. Therefore, the

probability or consequences of an event previously identified have

not been significantly changed.

(c) Allowing up to 12 hours to recover from the inoperability of

all three trains of Control Room Ventilation or two or more trains

of Fuel Handling Building HVAC does not represent a significant

change to the probability of an accident because the inoperability

of these ventilation systems are not identified as precursors to a

design basis event. The low likelihood of a design basis accident

during the limited period of allowed inoperability of these systems

does not represent a significant increase in the consequences of an

accident.

(d) The redefinition of plant operability requirements into

functional trains rather than individual components does not affect

the required system functional operability. Therefore, this change

does not represent an increase in the probability or consequences of

an accident previously identified.

(e) The alignment of the Control Room Ventilation System to the

same configuration it would be placed in from an actuation of the

inoperable radiation monitoring channel places the system in the

design condition. This alignment would result in maintaining the

control room envelope pressurized and increases the protection

afforded to the operators.

(f) The change in terminology does not change any requirements

or actions in the Specification. Therefore this change does not

represent an increase in the probability or consequences of any

accident previously evaluated.

Based on the above discussion, the individual changes do not

represent an increase in the probability or consequences of any

accident previously evaluated.

In addition to the changes proposed to controls over Control

Room Ventilation, Fuel Handling Building HVAC, and associated

actuation logic, an administrative change is proposed to remove the

footnote at the bottom of page 3/4 7-20. Since the footnote no

longer has meaning or relevance to the operation of the facility,

its removal does not increase the probability or consequences of any

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes make the existing Specifications internally

consistent, manually align a system to the actuated position,

provide an alternative measure that assures [that] a safety function

which is unavailable is not required to [be] perform[ed], provide an

extended period of allowance for all trains of a system to be

inoperable, and redefines system operability to reflect its

functional design. The proposed changes do not introduce any new

equipment into the plant or significantly alter the manner in which

existing equipment will be operated. The systems affected by the

proposed changes are not identified as contributing causal factors

in design basis accidents, their function is to assist in mitigation

of accidents postulated to occur. Since the proposed changes do not

allow activities that are significantly different from those

presently allowed, no possibility exists for a new or different kind

of accident from those previously evaluated.

In addition to the changes proposed to controls over reactivity

changes, an administrative change is proposed to remove the footnote

at the bottom of page 3/4 7-20. Since the footnote does not perform

any function and will never again apply to plant operations, its

removal cannot create the possibility of a new or different kind of

accident from those previously evaluated.

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

of safety?

The proposed changes do not involve a significant reduction in a

margin of safety because the ability of the Fuel Handling Building

HVAC and Control Room Ventilation Systems will be maintained. The

margin of safety is defined by the ability of the systems to limit

the release of radioactive materials and limit exposures to

operators respectively following a postulated design basis accident.

The only aspect of the proposed change that can be postulated to

have any effect on a margin of safety is the proposed allowance for

all trains of Control Room Ventilation or Fuel Handling Building

HVAC to be inoperable for a limited period. The low probability of a

design basis event that would require the system to perform its

safety function during the limited period allowed by the proposed

action assures that the change does not involve a significant change

in a margin of safety. Therefore, the proposed changes do not

significantly affect these operating restrictions and the margin of

safety which support the ability to make and maintain the reactor in

a safe shutdown and limit the release of radioactive material is not

affected.

In addition to the changes described above, an administrative

change is proposed to remove the footnote at the bottom of page

3/4 7-20. Since the footnote is no longer applicable to the

facility, its removal cannot result in a reduction in a margin of

safety.

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

this review, it appears that the standards of 10 CFR 50.92

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

Bockius, 1800 M Street, NW, Washington, DC 20036-5869.

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: September 29, 1998.

Description of amendment request: The licensee proposes to use a

revised methodology to calculate mass and energy release following a

postulated large-break loss-of-coolant accident. The amendment request

also included proposed changes to the 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, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or

[[Page 64124]]

consequences of an accident previously evaluated.

This proposal updates the design large break loss of coolant

accident (LBLOCA) analysis and methodology described in the UFSAR to

support replacement of Westinghouse Model E Original Steam

Generators (OSG) with Westinghouse Delta-94 Replacement Steam

Generators (RSG).

A safety analysis has been performed, including evaluation of

existing analyses and performance of bounding or confirming

calculations, to determine effects of the proposed changes.

Analysis of mass and energy releases and resultant containment

pressure and temperature response for the RSG concluded a small

reduction in peak pressure and temperature for the RSG compared to

the OSG. Thus, the proposed amendment does not involve a significant

increase in the probability of an accident previously evaluated.

Changes to the LBLOCA model caused by installation of the RSGs

and associated changes in analysis methodology result in no change

in radiological consequence as delineated in 10 CFR 100 and the

Standard Review Plan (NUREG-0800). Consequences of this design basis

accident have not increased.

Thus, changes in the LBLOCA design basis event analysis

associated with replacement of OSGs with RSGs do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This proposal updates the design basis large break loss of

coolant accident (LBLOCA) analysis and methodology described in the

Updated Final Safety Analysis Report (UFSAR) to support replacement

of OSGs with RSGs.

Fit, form, and design function of RSG equipment is not

significantly changed from OSG equipment. Analyses of LBLOCA mass

and energy releases and resultant containment system response

indicates that performance with RSGs remains within the existing

design limits. Thus, the proposed change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

A safety analysis has been performed, including evaluations of

existing analyses and performance of bounding and/or confirming

calculations, to determine the effect of the proposed changes.

Results of these analyses demonstrate that the proposed license

amendment and operation of STP Units with Delta-94 steam generators

installed will not produce post-accident Containment pressures or

temperatures exceeding existing Technical Specification limits.

Consequently, there are no effects on dose analyses due to design

basis LBLOCA performance of the RSGs. Radiological consequences of

the postulated accident did not change, and all results remain

within the acceptance criteria of 10 CFR 100 and the Standard Review

Plan (NUREG-0800).

Thus, the change in LBLOCA analysis results and methodology

descriptions in the UFSAR associated with replacement of Model E

steam generators with Delta-94 steam generators do not involve a

significant reduction in a margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

Bockius, 1800 M Street, NW, Washington, DC 20036-5869.

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: September 30, 1998.

Description of amendment request: The proposed amendment would

change the Updated Final Safety Analysis Report and revise the offsite

dose licensing basis to account for operation of the existing steam

generators at reduced feedwater inlet temperatures, and to account for

operation of the new replacement steam generators. The calculated

offsite dose consequences would increase for the main steamline break,

reactor coolant pump shaft seizure, and rod cluster control assembly

ejection accidents. The proposed increases in offsite doses are minimal

and all doses remain below the dose limits for their respective

accidents, as specified by 10 CFR Part 100 and the Standard Review Plan

(NUREG-0800).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This document updates the facilities' radiological design basis,

as described in the Updated Final Safety Analysis Report, to address

both a reduction in allowed nominal feedwater temperature for Model

E steam generators from 440 deg.F to 420 deg.F and the replacement

of Model E steam generators with Delta-94 steam generators.

Therefore, these changes do not change the probability of an

accident previously evaluated.

A safety analysis has been performed, including evaluations of

existing analyses and performance of bounding and/or confirming

calculations, to determine the impact of the proposed changes.

Effects on the dose analyses due to the accompanying physical

changes to the plant are slight. However, some improvements were

made to the analytical models used in the analyses. These

improvements were responsible for the majority of the increase in

offsite doses. While the radiological consequences of some

postulated accidents increased, all results remain within the

acceptance criteria, as defined in 10 CFR 100 and the Standard

Review Plan (NUREG-0800).

The radiological consequences of the postulated accidents remain

within their respective acceptance criteria with the use of the

revised analysis methodologies. Therefore, the change to allow

operation of the Model E steam generators at a reduced feedwater

temperature of 420 deg.F and the replacement of Model E steam

generators with Delta-94 steam generators do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This document updates the facilities' radiological design basis,

as described in the Updated Final Safety Analysis Report, to address

both a reduction in allowed nominal feedwater temperature for Model

E steam generators from 440 deg.F to 420 deg.F and the replacement

of Model E steam generators with Delta-94 steam generators. Since

the proposed changes to the Updated Final Safety Analysis Report are

analytical in nature, the changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

A safety analysis has been performed, including evaluations of

existing analyses and performance of bounding and/or confirming

calculations, to determine the impact of the proposed changes.

Effects on the dose analyses due to the accompanying physical

changes to the plant are slight. However, some improvements were

made to the analytical models used in the analyses. These

improvements were responsible for the majority of the increase in

offsite doses. While the radiological consequences of some

postulated accidents increased, all results remain within the

acceptance criteria, as delineated in 10 CFR 100 and the Standard

Review Plan (NUREG-0800), for the respective accidents.

The radiological consequences of the postulated accidents remain

within their respective acceptance criteria with the use of the

revised analysis methodologies. Therefore, the change to allow

operation of the Model E steam generators at a reduced feedwater

temperature of 420 deg.F and the replacement of Model E steam

generators with Delta-94 steam generators do not

[[Page 64125]]

involve a significant reduction in a margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: John N. Hannon.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: October 27, 1998.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Section 3/4.8.2.3, ``Electrical

Power Systems--DC Distribution--Operating,'' and the associated bases.

The surveillance requirements for battery testing would be revised.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the

proposed changes and determined that a significant hazards

consideration does not exist because operation of the Davis-Besse

Nuclear Power Station, Unit Number 1, in accordance with these changes

would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiators,

conditions, or assumptions are adversely affected by the proposed

changes to station battery testing methodology and frequency.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because no accident conditions or

assumptions are adversely affected by the proposed changes in

station battery testing methodology and frequency. The proposed

changes do not alter the source term, containment isolation, or

allowable radiological releases. The proposed changes are consistent

with the most recent IEEE Standard 450-1995, ``IEEE Recommended

Practice for Maintenance, Testing, and Replacement of Vented Lead-

Acid Batteries for Stationary Applications,'' and the ``Improved

Standard Technical Specifications for Babcock and Wilcox Plants,''

NUREG-1430, Revision 1.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

changes. The batteries are not an initiator or contributor to the

initiation of an accident. No new accident scenarios, transient

precursors, failure mechanisms, or limiting faults are introduced as

a result of the proposed changes.

3. Not involve a significant reduction in a margin of safety

because the proposed TS changes do not significantly reduce or

adversely affect the capabilities of any plant structures, systems

or components. These changes increase the effectiveness and

frequency of the battery tests being performed. Therefore, there is

not 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: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

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

NRC Project Director: Stuart A. Richards.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: October 27, 1998.

Description of amendment request: The proposed amendment would

relocate a Technical Specification (TS) surveillance requirement from

TS Section 3/4.6.5.1, ``Shield Building-Emergency Ventilation System''

to TS Section 3/4.6.5.2, ``Shield Building Integrity.'' Administrative

and bases changes would also be made.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

The Davis-Besse Nuclear Power Station has reviewed the proposed

changes and determined that a significant hazards consideration does

not exist because operation of the Davis-Besse Nuclear Power

Station, Unit Number 1, in accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiator is

affected by the proposed changes to the Technical Specifications

(TS) Index; TS Definition 1.6, ``Shield Building Integrity''; TS 3/

4.6.5.1, ``Emergency Ventilation System''; TS 3/4.6.5.2, ``Shield

Building Integrity''; TS Bases 3/4.6.5.1, ``Emergency Ventilation

System''; or TS Bases 3/4.6.5.2, ``Shield Building Integrity.''

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because no accident conditions and

assumptions are significantly affected by the above proposed

changes. The proposed change to relocate existing TS Surveillance

Requirement (SR) 4.6.5.1.d.4 to TS 3/4.6.5.2, and the subsequent

application of the Limiting Condition for Operation (LCO) of TS 3/

4.6.5.2 should the Emergency Ventilation System (EVS) be unable to

produce the required negative pressure in the annulus space due to

an opening in the ventilation boundary, would allow 24 hours to

restore the capability of maintaining the required negative pressure

in the annulus. The current SR 4.6.5.1.d.4 and associated TS LCO

3.6.5.1 would require entry into TS 3.0.3, thereby allowing only one

hour for restoration before commencing plant shutdown. The allowed

outage time of 24 hours is reasonable considering the limited

leakage design of containment and the low likelihood of a Design

Basis Accident (DBA) occurring during this time period. The proposed

changes are consistent with the guidance of the ``Improved Standard

Technical Specifications for Combustion Engineering Plants,'' NUREG-

1432, Revision 1 and the ``Improved Standard Technical

Specifications for Westinghouse Plants,'' NUREG-1431, Revision 1.

The ``Improved Standard Technical Specifications for Babcock and

Wilcox Plants,'' NUREG-1430, Revision 1 does not contain guidance

for shield building integri

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