Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 14, 1994, through November 25,

1994. The last biweekly notice was published on November 23, 1994 (59

FR 60377).

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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, 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 January 6, 1995, 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, Attention: Docketing and Services

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

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

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

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

Commission by a toll-free telephone call to Western Union at 1-(800)

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

should be given Datagram Identification Number N1023 and the following

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

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

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, 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.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: October 25, 1994

Description of amendments request: The proposed change would delete

the remainder of Appendix B, Environmental Technical Specifications,

including section 2/3.3.1, Water Level in the Discharge Canal, and

Section 2/3.4, Meteorology.

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:

Hydraulic - Water level in the Discharge Canal

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The deletion of the discharge canal

specification of 4.5 [plus or minus] 1 ft mean sea level (msl) with

daily monitoring still leaves the operating restrictions delineated

in Updated Final Safety Analysis Report (UFSAR) Section 2.4.8.3.3

(4.5 [plus or minus] 2 ft msl) and the National Pollutant Discharge

Elimination System (NPDES) permit requirements to minimize the

impact of the discharge canal on the local groundwater supply. As

stated in this UFSAR section, the effect on the local ground water

regime will be minimal within this band. Level recorders in the

control room facilitate the continued monitoring the discharge canal

level in excess of the Appendix B Environment Technical

Specification (ETS) listed daily surveillance. This change in no way

affects the design or operation of equipment that could initiate or

mitigate any accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. The discharge canal level is an environmental concern

with the effects of a spill of radioactive liquids (UFSAR Section

2.4.12.3) being a path to the intake canal, due to the areas natural

gradient. With the estimated travel time for the liquid to reach the

canal (intake) at 60 years, and the large flow rate the degree of

dilution is such that this does not pose a threat to local wells.

The amendment would not affect the operation or design of any plant

equipment; therefore, no new credible accidents are created. In

addition, the proposed amendment would not affect the capability of

the response systems to mitigate the consequences of any accident

previously evaluated; therefore, no new or different accident would

result from this change.

3. The proposed amendment does not involve a significant

reduction in a margin of safety. The existence of these ETS does not

provide a margin of safety related to the nuclear operation of the

site. No safety limits are affected by this change. Therefore, this

amendment would not result in a reduction in any margin of safety.

Meteorology

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The deletion of the meteorology specification

still leaves the program delineated in UFSAR Section 2.3.3, Onsite

Meteorological Measurements Program, and reporting/recording of the

hourly meteorological data required to support Technical

Specification 6.9.1.10.a, Semiannual Radioactive Effluent Release

Report. This program is based on the meteorological monitoring

program described in Regulatory Guide 1.23, and NUREG-0654. While

the existing ETS 30 day reporting requirement for extended out-of-

service time and shiftly manual acquisition of data during batch or

accidental releases are not otherwise covered, the program does

contain the Regulatory Guide 1.23 reference to 90% data recovery and

a backup phone line is available for data retrieval. This proposed

amendment in no way affects the design or operation of equipment

that could initiate or mitigate any accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. The meteorological monitoring program is specified in

UFSAR Section 2.3.3. The amendment would not affect the operation or

design of any plant equipment; therefore, no new credible accidents

are created. In addition, the proposed amendment would not affect

the capability of the response systems to mitigate the consequences

of any accident previously evaluated nor would the amendment reduce

the effectiveness of the Emergency Response Plan; therefore, no new

or different accident would result from this change.

3. The proposed amendment does not involve a significant

reduction in a margin of safety. The meteorological ETS does not

provide any additional margin of safety related to the operation of

a nuclear plant. The meteorological program established in the UFSAR

covers the requirements stated in Appendix E to 10 CFR 50 by

providing meteorological systems adequate for determining the

magnitude of, and for continuously assessing the impact of, the

release of radioactive materials to the environment. The

meteorological instrumentation is used to measure environmental

parameters which may affect distribution of fission products and

gases following a Design Basis Accident (DBA); however, it is not a

primary success path for the mitigation of a DBA. No safety limits

are affected by this change. Therefore, this amendment would not

result in a reduction in any margin of safety.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: William H. Bateman

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: October 28, 1994

Description of amendments request: The proposed changes would

revise the Technical Specifications to increase the surveillance test

intervals and allowable out-of-service times for selected

instrumentation addressed in Section 3/4.3. The proposed changes would

permit specified channel functional tests to be conducted quarterly

rather than weekly or monthly. Specifically, the proposed changes would

revise the surveillance test intervals and allowable out-of-service

times for the reactor protection system instrumentation, isolation

actuation instrumentation, emergency core cooling system actuation

instrumentation, control rod withdrawal block instumentation, control

room emergency ventilation system instrumentation, anticipated

transient without scram - recirculation pump trip system

instrumentation, end-of-cycle recirculation pump trip system

instrumentation, and reactor core isolation cooling system actuation

instrumentation, in accordance with NRC-approved General Electric

Company Licensing Topical Reports and NUREG-1433, Standard Technical

Specifications, General Electric Plants, BWR/4.''

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 Brunswick Steam Electric Plant, Units 1 and 2,

in accordance with the proposed amendment, would not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The generic Licensing Topical Report, NEDC-30851P-A, assessed

the impact of changing RPS surveillance test intervals (STIs) and

allowable out-of-service times (AOTs) on the RPS failure frequency,

the scram frequency and equipment cycling. Specifically, Section

5.7.4, ``Significant Hazards Assessment'' of NEDC-30851P-1 states:

``Fewer challenges to the safeguards system, due to less

frequent testing of the RPS, conservatively results in a decrease of

approximately one percent in core damage frequency''. This decrease

is based upon the following:

* Based on the plant-specific experience presented in Appendix

J, the estimated reduction in scram frequency (0.3 scrams/yr)

represents a 1 to 2 percent decrease in core damage frequency based

on the BWR plant specific Probabilistic Risk Assessments (PRAs)

listed in Table 5-8.

* The increase in core damage frequency due to less frequent

testing is less than one percent. This increase is even lower (less

than 0.01 percent) when the changes resulting from the

implementation of the Anticipated Transients Without Scram (ATWS)

rule are considered. Therefore, this increase is more than offset by

the decrease in CDF due to fewer scrams.

* The effect of reducing unnecessary cycles on RPS equipment,

although not easily quantifiable also results in a decrease in core

damage frequency.

* The overall impact on core damage frequency of the changes in

allowable out-of-service time is negligible.''

From this generic analysis, the BWR Owners' Group concluded and

CP&L concurs that the proposed changes do not significantly increase

the probability or consequences of an accident previously evaluated,

since the increase in probability of a a scram failure due to RPS

unavailablity is insignificant. The overall probability of an

accident is decreased as the time the RPS instrumentation logic

operates undisturbed is increased, resulting in fewer inadvertent

scrams during testing and repair. The proprietary plant-specific

analysis contained in this submittal (Enclosure 6) demonstrates

that, although BSEP Units 1 and 2 differ from the generic plant

analyzed in LTR NEDC-30851P-A, the net effect of the plant-specific

differences does not alter the generic conclusions.

The generic Licensing Topical Reports, NEDC-30851P-A, Supplement

2 and NEDC 31677P-A, assessed the impact of changing STIs and AOTs

for BWR Isolation Instrumentation. Section 4.0, ``Summary of

Results,'' of NEDC-30851P-A, Supplement 2 states:

``The results indicate that the effects on probability of

failure to initiate isolation are very small and the effects on

probability or frequency of failure to isolate are negligible in

nearly every case. In addition, the results indicate that increasing

the AOT to 24 hours for tests and repairs has a negligible effect on

the probability of failure of the isolation function. These combined

with changes to the testing intervals and allowable out-of service

times for RPS and ECCS instrumentation provide a net improvement to

plant safety and operations.''

and Section 5.6, ``Assessment of Net Effect of Changes,'' of

NEDC-31677P-A states:

``A reduction in core damage frequency (CDF) of at least as much

as estimated in the ECCS instrumentation analysis can be expected

when the isolation actuation instrumentation STIs are changed from

one month to three months. The chief contributor to this reduction

is the channel functional tests for the MSIVs. Inadvertent closure

of the MSIVs will cause an unnecessary plant scram. This reduction

in CDF more than compensates for any small incremental increase (10%

or 1.0E-07/year) in calculated isolation function failure frequency

when the STI is extended to three months.''

From this generic analysis, the BWR Owners' Group concluded and

CP&L concurs that the proposed changes do not significantly increase

the consequences of an accident previously evaluated, since the

increase in probability of an isolation failure due to isolation

instrumentation unavailability is insignificant. For those

parameters common to RPS, the overall probability of an accident is

actually decreased as the time the RPS instrumentation logic

operates undisturbed is increased, resulting in less inadvertent

scrams during testing and repair. The plant-specific evaluation

provided with this submittal (Enclosure 8) demonstrates that the

conclusions of the generic analyses are applicable to BSEP Units 1

and 2.

The generic Licensing Topical Report, NEDC-30936P-A (Parts 1 and

2), assessed the impact of changing STIs and AOTs for all BWR ECCS

Actuation Instrumentation. Section 4.0, ``Technical Assessment of

Changes,'' of NEDC-30936P-A (Part 2) states:

``The results indicate an insignificant (less than 5E-7 per

year) increase in water injection function failure frequency when

STIs are increased from 31 days to 92 days, AOTs for repair of the

ECCS actuation instrumentation are increased from one hour to 24

hours, and AOTs for surveillance testing are increased from two to

six hours. For all four BWR models the increase represents less than

4% increase in failure frequency. However, when other factors which

influence the overall plant safety are considered, the net result is

judged to be an improvement in plant safety.''

From this generic analysis, the BWR Owners' Group concluded and

CP&L concurs that the proposed changes do not significantly increase

the probability or consequences of an accident previously evaluated,

since the increase in probability of a water injection failure due

to ECCS instrumentation unavailability is insignificant and the net

result is judged to be an improvement in plant safety. The plant-

specific analysis contained in this submittal (Enclosure 7)

demonstrates that, although BSEP Units 1 and 2 differ from the

generic model analyzed in LTR NEDC-30936P-A, the net effect of the

plant-specific differences does not alter the generic conclusions.

The generic Licensing Topical Reports, NEDC-30851P-A, Supplement 1,

and GENE-770-06-1-A assessed the impact of changing Control Rod

Block STIs and AOTs on Rod Block failure frequency. GENE-770-06-1-A

also assessed the impact of changing STIs and AOTs on ATWS-RPT and

EOC-RPT failure frequency. Section 5 (Brookhaven National

Laboratory's Technical Evaluation Report - Attachment 2 to the NRC

SER) of NEDC-30851P-A, Supplement 1 states:

1``The BWR Owners' Group proposed changes to the Technical

Specifications concerning the test requirements for BWR control rod

block instrumentation. The changes consist of increasing the

surveillance test intervals from one to three months. These test

interval extensions are consistent with the already approved changes

to STIs for the Reactor Protection System. The technical analysis

reviewed and verified as documented herein indicates that there will

be no significant changes in the availability of the control rod

block function if these changes are implemented. In addition, there

will be a negligible impact on the plant core melt frequency due to

the decreased testing.''

and Section 2.0, ``Summary'' of GENE-770-06-1-A states:

``Technical bases are provided for selected proposed changes to

the instrumentation STIs and AOTs that were identified in the BWROG

Improved BWR Technical Specification activity. These STI and AOT

changes are consistent with approved changes to the RPS, ECCS, and

isolation actuation instrumentation. These proposed changes do not

result in a degradation to overall plant safety.''

Based on the generic analysis in NEDC-30851P-A, Supplement 1,

the BWR Owners' Group concluded and CP&L concurs that the proposed

changes to Control Rod Withdrawal Block instrumentation do not

significantly increase the probability or consequences of an

accident previously evaluated. Also, based on the generic assessment

in GENE-770-06-1-A, the BWR Owners' Group concluded and CP&L concurs

that the proposed changes to the ATWS-RPT and EOC-RPT

instrumentation do not significantly increase the probability or

consequences of an accident previously evaluated.

Bases contained in GE Topical Report GENE-770-06-2P-A, assessed

the impact of changing STIs and AOTs on BWR RCIC failure frequency.

Section 2.0, ``Summary'' of GENE-770-06-2P-A states:

``The STI and AOT changes to the RCIC actuation instrumentation

are justified based on their small effect on the water injection

function unavailability and consistency with comparable changes to

actuation instrumentation for the other ECCS subsystems.''

On this basis, the BWR Owners' Group concluded and CP&L concurs

that the proposed changes to RCIC instrumentation do not

significantly increase the probability or consequences of an

accident previously evaluated.

2) Operation of Brunswick Steam Electric Plant, Units 1 and 2,

in accordance with the proposed amendment, would not create the

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

previously evaluated.

The proposed changes do not alter the physical characteristics

or function of any plant systems or components and they do not

introduce any new mode of operation. Therefore, system and component

performance would not be challenged in a manner that could create

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

accident previously evaluated.

3) Operation of Brunswick Steam Electric Plant, Units 1 and 2,

in accordance with the proposed amendment, would not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed and approved the generic studies

contained in the LTRs and has concurred with the BWR Owners' Group

that the proposed changes do not significantly affect the

probability of failure or availability of the affected Instrument

Systems. The proposed changes to AOTs provide realistic times to

complete the required actions without increasing the overall

instrument failure frequency. Likewise, the extended STIs do not

result in significant changes in the probability of instrument

failure. Furthermore, the proposed changes will reduce the

probability of test-induced plant transients and equipment failures.

Finally, instrument setpoint drift will remain within present

tolerances, thereby assuring that the margin of safety, as

demonstrated by applicable safety analyses, remains unchanged.

Therefore, it is concluded that the proposed changes would not

result in a 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: William H. Bateman

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: November 4, 1994

Description of amendment request: The requested amendment will

change the testing frequency of the turbine overspeed protection valves

from monthly to quarterly.

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 requested change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The requested change will have no influence on the

probability or consequences of an accident previously evaluated. The

accident of concern to this requested change is a turbine overspeed

with missile generation impacting safety related components or

structures. The evaluation in WCAP-11525 shows that the probability

of a missile ejection incident will not be affected with the

requested frequency reduction to the turbine overspeed protection

valve surveillance test. There is no change to the consequences of

the event as the postulated accident event is unchanged.

Accordingly, the requested change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated. The change affects the test interval for the turbine

overspeed protection valves and does not change the design,

operation, or failure modes of the valves and other components in

the turbine overspeed protection system. Therefore, the requested

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

accident from any accident previously evaluated.

3. The requested change does not involve a significant reduction

in the margin of safety. The probability of turbine overspeed with

an extension of the testing interval has been determined to be

within applicable acceptance criteria. The change does not affect

the design, operation, or failure modes of the valves or other

components in the turbine overspeed protection system. Accordingly,

the requested change will not involve a significant reduction in the

margin of safety.

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

this review, it appears that the three standards of 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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: William H. Bateman

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

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

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

Will County, Illinois

Date of amendment request: November 7, 1994

Description of amendment request: The proposed amendment would

allow an increase to the allowable nominal fuel enrichment from 4.2 to

5.0 weight percent Uranium-235 (w/o U-235). The changes include: (1)

increasing the allowable storage enrichment in Region 1 and allowing

the use of Integral Fuel Burnable Absorbers (IFBAs) for reactivity

equivalencing, (2) revising the Region 2 discharge burnup curve to

include nominal fuel enrichments up to 5.0 w/o U-235, and (3) making

editorial changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes to Section 5 of Technical Specifications do

not affect any accident initiators or precursors and do not change

or alter the design assumptions for the systems or components used

to mitigate the consequences of an accident. The fuel enrichment

increase will not affect reactor operation or the core design

methods. The physical characteristics of the fuel assemblies are not

changed, and fuel assembly movement will continue to be controlled

by approved fuel handling procedures. Reload core designs will

continue to be performed on a cycle by cycle bases as part of the

reload safety evaluation process, using NRC approved codes and

methods. Each reload design is evaluated to confirm that the cycle

core design adheres to the limits that exist in the accident

analyses and Technical Specifications to ensure that reactor

operation is acceptable.

The proposed changes are consistent with the analysis performed

in the ``Criticality Analysis of Byron and Braidwood Station Fuel

Storage Racks.'' The analysis was revised in June 1994 to include

boraflex gaps and shrinkage. The revised analysis is provided in the

proposed Technical Specification amendment. The analysis methodology

has been previously accepted by the NRC and is consistent with the

appropriate standards to establish the Keff limit for storage

racks and to calculate the maximum Keff. The reanalysis

addresses the most limiting postulated accident of a misloaded fuel

assembly and has shown that having at least 300 ppm of soluble boron

offsets any positive reactivity impacts for any of the postulated

accidents. The concentration of boron in the spent fuel pool water,

which is administratively controlled, is sufficient to maintain

Keff less than or equal to 0.95. The analysis is bounding for a

dropped fuel assembly on top of a rack or between rack modules, loss

of cooling systems, and reduction the fuel pool temperature to less

than 50 deg.F. The proposed changes do not impact any other accident

previously evaluated in the [Updated Final Safety Analysis Report]

UFSAR. There is no postulated accident that could cause reactivity

to increase beyond the analyzed conditions in the spent fuel racks.

There is no impact on the ability of the Spent Fuel Pool cooling

system to maintain the bulk pool temperature within limits. The

UFSAR analysis performed to calculate the maximum fuel cladding

temperature and spent fuel pool cooling include assumptions which

bound the use of more highly enriched fuel assemblies. Although fuel

enrichment is not a specific assumption in any of these analyses,

the heat load of a typical core offload may change with higher

enrichments. The average burnup of the offload will be increased

since few assemblies will be used per cycle; however, the new heat

load will continue to be [bound] by the UFSAR analysis because the

spent fuel pool racks have been analyzed for a total core offload

with all fuel assemblies having 4.5 years of operating time.

The radiological consequences analysis continues to bound the

licensed fuel burnup and enrichment at Byron and Braidwood stations.

The radiological consequences analysis results are a function of the

core inventory of radioactive isotopes. Since the maximum fuel

burnup limits and fuel peaking factors will not be exceeded, the

assumed fission product inventory will remain valid; therefore, the

limits of 10 CFR [Part] 100 continue to be met. Additionally, Byron

and Braidwood addressed the issue of the impact on the radiation

levels at the pool surface to the worker during non-accident

conditions. These conditions are not changed as [a] result of this

submittal, because the average fuel assembly burnup limit (isotopic

inventory) and maximum power produced in each fuel assembly will not

be changed by the increased fuel enrichment.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not affect the design or operation of

any system, structure, or component in the plant. There are no

changes to parameters governing plant operation; no new or different

type of equipment will be installed. Each reactor core design will

continue to meet all design requirements; operation of the core will

not be affected. No modifications to the spent fuel pool are being

pursued and the fuel parameters used in the analysis remain

bounding. The method and manner in which the fuel will be stored in

the spent fuel pool has not changed. The proposed changes ensure

that 17X17 (Optimized Fuel Assembly, VANTAGE 5, VANTAGE +, and

PERFORMANCE +) fuel assemblies can be safely stored, maintaining a

Keff less than or equal to 0.95 under full water density

conditions, in both Regions 1 and 2 of the spent fuel pool. All

design criteria and criticality acceptance criteria continue to be

met. The reanalysis addresses the most limiting postulated accident

(misloaded fuel assembly) and has shown that having at least 300 ppm

of soluble boron offsets any positive reactivity impacts for any of

the postulated accidents. The level of boron in the spent fuel pool

water, which is administratively controlled, is sufficient to

maintain Keff less than or equal to 0.95. The reanalysis to

increase the storage enrichment of fuel in Regions 1 and 2 of the

spent fuel pool does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Additionally, approval of this amendment will not create a new

accident with regards to the new fuel storage vault which is

designed to handle the increased enrichment. The Byron and Braidwood

new fuel vaults were previously analyzed using NRC accepted

criticality analysis methodology in June 1989. This analysis was

performed to increase the storage enrichment of the New Fuel Vault

to 5.0 w/o U-235. The New Fuel Vault analysis was submitted to the

NRC and is the current licensing basis.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the margin of safety for any

Technical Specification. All reactor design criteria will continue

to be met. The methodologies used in the accident analyses have been

accepted previously by the NRC and all criticality acceptance

criteria have been met under all assumed conditions (normal and

accident). The design basis for preventing criticality outside the

reactor is that, including uncertainties, there is a 95 percent

probability at a 95 percent confidence level that the effective

neutron multiplication factor, Keff, of the fuel assembly array

will be less than 0.95 as recommended by ANSI 57.2-1983 and OT

Position Paper for Review and Acceptance of Spent Fuel Storage and

Handling Applications, dated April 14, 1978. The analyses for both

Regions 1 and 2 fuel storage were verified to meet the above design

basis.

The criticality analysis for Regions 1 and 2 has been revised to

allow for storage of fuel assemblies with enrichments up to 5.0 w/o

U-235. The proposed Technical Specification changes include those

changes necessary to maintain Keff less than or equal to 0.95,

including conservative allowances for uncertainties and biases, when

the pool is flooded with unborated water. The proposed changes

include a requirement for fuel assemblies with enrichments above 4.2

w/o U-235 to contain sufficient integral fuel burnable absorbers

such that the maximum reference fuel K infinity is less than or

equal to 1.470 in unborated water at 68 deg.F due to restrictions on

spent fuel storage. Should a postulated accident occur which causes

a reactivity increase in the Byron and Braidwood Spent Fuel Pools,

Keff will be maintained less than or equal to 0.95 due to the

presence of at least 300 ppm of soluble boron in the spent fuel

pool. The proposed changes do not affect any plant safety parameters

or setpoints.

The proposed changes ensure that the design basis for preventing

criticality in the fuel storage areas is preserved, and fuel cycle

designs will continue to be analyzed using NRC accepted codes and

methods to ensure the design bases are satisfied.

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: For Byron, the Byron Public

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company (CYAPCO), and Northeast

Nuclear Energy Company (NNECO), Docket Nos. 50-213 and 50-245,

Haddam Neck Plant, and Millstone Nuclear Power Station, Unit 1,

Middlesex County, and New London County, Connecticut

Date of amendment request: October 31, 1994

Description of amendment request: The proposed amendments would

renew the existing license conditions for both plants to implement and

maintain Integrated Implementation Schedule (IIS) Program Plans (the

Program Plan). The Program Plans provide a methodology to be followed

for scheduling plant modifications and engineering evaluations.

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:

YAPCO and NNECO have reviewed the proposed changes in accordance

with 10 CFR 50.92 and conclude that the changes do not involve a SHC

[significant hazards consideration]. The basis for this conclusion

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

The proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Operation of the facilities in accordance with these proposed

changes would require the implementation of the IIS methodology

described in the Program Plans. As such, it requires that CYAPCO and

NNECO establish an administrative means for tracking, prioritizing,

and scheduling NRC-required plant modifications and engineering

evaluations, and licensee identified plant improvement projects.

This methodology is intended to enhance plant safety by more

effectively controlling the number and scheduling of plant

modifications, thereby assuring that issues required for safe

operation of the plants receive priority and are completed in a

timely manner. Because the license conditions address only an

administrative scheduling mechanism, it does not affect directly the

design or operation of the plant. Therefore, no accident analyses

are affected and the proposed changes do not increase the

probability or consequences of any previously evaluated accident.

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

from any previously analyzed.

The proposed license conditions establish a requirement related

to scheduling of modifications and engineering evaluations. Because

the license conditions address only an administrative scheduling

mechanism, they do not affect directly the design or operation of

the plants. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from those

previously evaluated.

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

The proposed license conditions renew administrative

requirements intended to enhance public safety and reliable plant

operation. The proposed license conditions do not affect any

accident analyses, directly modify the plant configurations, or

change the way the plants are operated. The methodologies are

intended to enhance plant safety by more effectively controlling the

number and scheduling of plant modifications, thereby assuring that

issues required for safe operation of the plants receive priority

and are completed in a timely manner. Because the license conditions

address only an administrative scheduling mechanism, they do not

affect directly the design or operation of the plants. Therefore,

the proposed changes do not involve a reduction in any 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: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457, for the Haddam Neck Plant, and

the Learning Resource Center, Three Rivers Community-Technical College,

Thames Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for

Millstone Unit 1.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee

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

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

Date of amendment request: August 5, 1994, as supplemented on

November 17, 1994.

Description of amendment request: This amendment is an additional

followup to the amendment request of May 29, 1992, published in the

Federal Register on July 8, 1992 (57 FR 30242), which changed the

Technical Specifications (TSs) Section 1.0, Definitions, to accommodate

a 24-month fuel cycle and which proposed the extension of the test

intervals for specific surveillance tests. This amendment proposes

extending the surveillance intervals to 24 months for the following

additional surveillance tests:

(1) Charging Flow Instrumentation

(2) Containment Sump, Recirculation Sump, and Reactor Cavity

Continuous Level Instrument Channels

(3) Auxiliary Feedwater Flow Rate Channel

(4) Control Room Air Filtration System

(5) Post Accident Containment Venting System

(6) Liquid Rad-Waste Flow Channel

(7) Steam Generator Blowdown Flow Channel

(8) Liquid Waste Distillate Tank Level Channels

(9) Primary Water Storage Tank Level Instrumentation

(10) Flow Rate Monitors; Plant Vent (Unit 2) and Stack Vent (Unit

1)

(11) Stack Vent Noble Gas Activity Monitor (R-60)

(12) High Pressure Water Fire Protection System

(13) Fire Protection System Diesel Engine

(14) Electrical Tunnel, Diesel Generator Building, and Containment

Fan Cooler Fire Protection Spray Systems; (A) System Functional Test

and (B) Spray Header Visual Inspection

(15) Penetration Fire Barriers

(16) Smoke Detectors/Electrical Penetration Area Inside Containment

(17) Functional Testing of Containment Sump Pumps

The changes requested by the licensee are in accordance with

Generic Letter 91-04, ``Changes in Technical Specification Surveillance

Intervals to Accommodate a 24-Month Fuel Cycle.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Charging Flow Instrumentation

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

Charging Flow instrumentation be changed from 18 months (+25%) to

every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

channel statistical error resulting from a 30 month operating cycle.

The existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin. Plant

equipment will provide protective functions to assure that Safety

Analysis limits are not exceeded. This will prevent the possibility

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

from occurring.

3. A significant reduction in a margin of safety is not

involved.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds current margin. This margin, which is

equivalent to the existing margin, is necessary to assure that

protective safety functions will occur so that Safety Analysis

limits are not exceeded.(2) Containment Sump, Recirculation Sump,

and Reactor Cavity Continuous Level Instrument Channels

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the calibration and test frequency for the

Containment Sump, Recirculation Sump and Reactor Cavity continuous

level monitoring instrument channels be revised from every 18 months

(+25%) to 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

channel statistical error resulting from a 30 month operating cycle.

The existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds current margin. Plant equipment

will provide protective functions to assure that Safety Analysis

limits are not exceeded. This will prevent the possibility of a new

or different kind of accident from any previously evaluated from

occurring.

3. There has been no reduction in the margin of safety.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds current margin. This margin is necessary to

assure that protective safety functions will occur so that safety

analysis limits are not exceeded.

(3) Auxiliary Feedwater Flow Rate Channel

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the calibration frequency for the Auxiliary

Feedwater Flow Rate channel be revised from 18 months (+25%) to 24

months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists between the existing

technical specification limits and the licensing basis Safety

Analysis limit to accommodate the channel statistical error

resulting from a 30 month operating cycle. The existing margin

between the Technical Specification limit and the Safety Analysis

limit provides assurance that plant protective actions will occur as

required. It is therefore concluded that changing the surveillance

interval from 18 months (+25%) to 24 months (+25%) will not result

in a significant increase in the probability or consequences of an

accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin between the

existing Technical Specification limit and the Safety Analysis

limit. Plant equipment, which will be set at (or more conservatively

than) Technical Specification limits, will provide protective

functions to assure that safety analysis limits are not exceeded.

This will prevent the possibility of a new or different kind of

accident from any previously evaluated from occurring.

3. There has been no reduction in the margin of safety.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the margin which exists between the current

Technical Specification limit and the licensing basis Safety

Analysis limit. This margin, which is equivalent to the existing

margin, is necessary to assure that protective safety functions will

occur so that Safety Analysis limits are not exceeded.

(4) Control Room Air Filtration System

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the surveillance frequency for the Control

Room Air Filtration System be changed from every 18 months (+25%) to

every 24 months (+25%).

For the flow tests, data from 1986 to date indicates that the

Control Room Filtration System performed in an acceptable manner

when surveilled on an 18 month (+25%) basis. The only discrepancy

was due to a hardware error and was independent of the time between

surveillances. Per Generic Letter 91-04, this past test history

provides an adequate basis to conclude that an extended operating

cycle would have minimal impact upon the flow characteristics of the

Control Room Filtration System. The modification of the filtration

system in 1993 only enhanced system performance.

With regard to the absorbance properties of the charcoal,

previous test data highlights a problem occurring during the 1986-

1987 period which subsequent testing confirms was adequately

resolved.

With the 1993 modification which increased the carbon bed

thickness from 1'' to 4'', performance can only be enhanced.

Therefore, it is concluded that a significant increase in the

probability or consequences of an accident previously evaluated will

not be incurred by changing the surveillance interval from 18 months

(+25%) to 24 months (+25%).

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

A review of past historical surveillance data over 7 years

indicates no failures which were time dependent. The modification,

which was performed in 1993, can only enhance performance of the

system. New fans, an increased charcoal bed thickness, and new HEPA

[high-efficiency particulate air] filters will increase the

reliability of the system. Thus, it is concluded that the

possibility of a new or different kind of accident than that

previously evaluated has not been created.

3. There has been no significant reduction in the margin of

safety.

Past test data validated the acceptability of the previous air

filtration system for an extended surveillance interval. The

modification performed in 1993 will only enhance the reliability and

performance of the air filtration system. Thus, it is concluded that

a significant reduction in the margin of safety is not involved.

(5) Post Accident Containment Venting System

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the surveillance frequency for the Post

Accident Containment Venting system be revised from every 18 months

(+25%) to 24 months (+25%).

A review of past test history from 1986 to date indicates that

the Post Accident Containment Venting System performed in a

satisfactory manner when the surveillance period was 18 months

(+25%). There was one discrepant condition noted in the 1989 test,

which, based upon subsequent tests in 1991 and 1993, does not appear

to have been age related. The 1989 observation concerning a gasket

is considered to be a one time only event and unlikely to reoccur as

a result of extending the surveillance interval from 18 months

(+25%) to 24 (+25%).

An added consideration, in terms of safety significance, is the

fact that the Post Accident Containment Venting system is diverse

and redundant to the post accident hydrogen recombiners which are

themselves redundant and the primary means of reducing the post

accident hydrogen concentration within containment. The venting

system is not relied upon for containment pressure control.

Due to the satisfactory past test history of the venting system,

together with its secondary role as a means of controlling post

accident hydrogen concentration, it is concluded that a significant

increase in the probability or consequences of an accident

previously evaluated will not be incurred by changing the

surveillance interval from 18 months (+25%) to 24 months (+25%).

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

A review of past historical surveillance data over 7 years

indicates no failures which are considered to be time dependent.

Although one discrepant condition was observed in the 1989 test it

was not repeated in subsequent surveillances. Per Generic Letter 91-

04, this constitutes a sufficient basis for revising the

surveillance interval from 18 months (+25) to 24 months (+25%). This

extension in the operating interval is not expected to have an

impact upon the availability of the system. Thus, it is concluded

that the possibility of a new or different kind of accident

previously evaluated has not been created.

3. There has been no reduction in the margin of safety.

As past test data validates the presumption that an extended

operating cycle will not impact the availability of the Post

Accident Containment Venting Systems, it is concluded that a

significant reduction in the margin of safety is not involved.

(6) Liquid Rad-Waste Flow Channel

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Liquid Rad-Waste Flow Channel be revised from every 18 months

(+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

channel statistical error resulting from a 30 month operating cycle.

The existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds current margin. Plant equipment

will be set to provide protective functions to assure that Safety

Analysis limits are not exceeded. This will prevent the possibility

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

from occurring.

3. There has been no reduction in the margin of safety.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the allowable operating margin. This margin,

which is equivalent to the existing margin, is necessary to assure

that protective safety functions will occur so that Safety Analysis

limits are not exceeded.

(7) Steam Generator Blowdown Flow Channel

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Steam Generator Blowdown Flow channel be revised from every 18

months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

channel statistical error resulting from a 30 month operating cycle.

The existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin. Plant

equipment will provide protective functions to assure that Safety

Analysis limits are not exceeded. This will prevent the possibility

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

from occurring.

3. There has been no reduction in the margin of safety.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the allowable operating margin. This margin,

which is equivalent to the existing margin, is necessary to assure

that protective safety functions will occur so that Safety Analysis

limits are not exceeded.

(8) Liquid Waste Distillate Tank Level Channels

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Liquid Waste Distillate Tank level of tanks 13 and 14 be revised

from every 18 months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

channel statistical error resulting from a 30 month operating cycle.

The existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds current margin. Plant equipment

will be set to provided protective functions to assure that Safety

Analysis limits are not exceeded. This will prevent the possibility

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

from occurring.

3. There has been no reduction in the margin of safety.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the allowable operating margin. This margin,

which is equivalent to the existing margin, is necessary to assure

that protective safety functions will occur so that safety analysis

limits are not exceeded.

(9) Primary Water Storage Tank Level Instrumentation

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Primary Water Storage Tank Level instrumentation be changed from

every 18 months (+25%) to 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

surveillance has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

channel statistical error resulting form a 30 month surveillance.

The existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in surveillance interval will result in a

channel statistical allowance which can be accommodated over a 30

month operating cycle. Plant equipment, which will be set at (or

more conservatively than) Technical Specification limits, will

provide protective functions to assure that Safety Analysis limits

are not exceeded. This will prevent the possibility of a new or

different kind of accident from any previously evaluated from

occurring.

3. There has been no significant reduction in the margin of

safety.

The above changes in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds current margin. This margin is necessary to

assure that protective safety functions will occur so that Safety

Analysis limits are not exceeded.

(10) Flow Rate Monitors; Plant Vent (Unit 2) and Stack Vent

(Unit 1)

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the calibration frequency for the flow rate

monitors for the Plant Vent (Unit 2) and the Stack Vent (Unit 1) be

revised from every 18 months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists to accommodate the

statistical error resulting from a 30 month operating cycle. The

existing margin provides assurance that plant protective actions

will occur as required. It is therefore concluded that changing the

surveillance interval from 18 months (+25%) to 24 months (+25%) will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a statistical

allowance which exceeds the current margin. Plant equipment will be

calibrated to provide data to assure that safety analysis limits are

not exceeded. This will prevent the possibility of a new or

different kind of accident from an previously evaluated from

occurring.

3. There has been no reduction in the margin of safety.

The proposed change in the surveillance interval resulting from

an increased operating cycle will not result in a channel

statistical allowance which exceeds the allowable operating margin.

This margin, which is equivalent to the existing margin, is

necessary to assure that protective safety functions will occur so

that safety analysis limits are not exceeded.

(11) Stack Vent Noble Gas Activity Monitor (R-60)

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the calibration frequency for the stack vent

noble gas activity monitor be revised from every 18 months (+25%) to

every 24 months (+25%).

The current monitor replaced the previous monitor and therefore

there is only one refueling cycle surveillance data available which

proved to be satisfactory. The vendor recommends a calibration

period based on user experience. Insofar as the 18 month (+25%)

surveillance has proven to be acceptable, extension to a 24 month

(+25%) cycle is consistent with the vendor's recommendation. Any

additional uncertainty generated due to the extended surveillance is

bounded by the uncertainty inherent in a grab sample taken once per

24 hours which is the required compensatory action should the

monitor be inoperable. Since setpoints for alarms are not critical

to either plant operation or safety, since extensive margin is

reflected between the setpoint and applicable limits, it is

concluded that any additional uncertainty involved in a longer

surveillance cycle will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

This monitor measures the activity of potentially radioactive

gaseous effluent through the stack vent. The alarm setpoints are set

at a point sufficiently above expected radioactivity levels to avoid

unnecessary alarms and, at the same time, far below discharge

limits. The purpose of the monitor is to annunciate in the event an

unexpected spike in radioactivity level should occur so that

corrective action can be taken prior to exceeding a discharge limit.

The margin that exists between the discharge limit and the setpoint

is more than sufficient to accommodate any drift that could be

practically expected in a 24 month (+25%) operating cycle.

In this capacity, the monitor does not have setpoints which are

critical to plant operation or safety. Readings are not used in a

quantitative manner nor is accuracy important. It is important that

the instrument remain operable and respond to step changes in

radioactivity level over the operating cycle. It is therefore

concluded that an extended operating cycle will not result in the

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

previously evaluated.

3. There has been no reduction in the margin of safety.

Sufficient margin exists between plant setpoints and applicable

limits to accommodate any realistic drift projected to occur over a

30 month operating cycle. Furthermore, instrument indications are

not used in a quantitative manner nor is instrument accuracy of

importance. Therefore, it is concluded that no significant reduction

in the margin of safety will result from an extended operating

cycle.

(12) High Pressure Water Fire Protection System

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the system functional test of the High-

pressure Water Fire Protection System be changed from every 18

months (+25%) to every 24 months (+25%).

This system is a static system which is not normally required to

operate. The main fire pumps are on standby and are not in operation

except for testing. Thus, almost no wear is induced as a function of

time except that which results from being in standby status which is

minimal and slow acting. Under these circumstances, extending the

operating cycle between surveillances would be expected to have

negligible affect upon system operability. It is therefore concluded

that there would be no significant increase in the probability or

consequences of an accident as a result of an extended interval

between surveillances.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Extension of the plant operating cycle will primarily extend the

time the pumps are in standby capacity. The potential for system

deterioration is minimal under these circumstances. Any

deterioration that does occur will be slow acting with respect to

time. A significant deterioration would be detected by a monthly

pump operating test. Thus, an extended operating cycle is not

expected to create the possibility of a new or different kind of

accident form [from] any previously analyzed.

3. There has been no reduction in the margin of safety.

Extension of the operating cycle by several months only serves

to extend the period of time when the pumps are in standby status.

Any deterioration under these circumstances will be slow acting.

Significant deterioration would be detected by the monthly operating

test. Therefore, it is concluded that an extended interval between

surveillances will involve no significant reduction in the margin of

safety.

(13) Fire Protection System Diesel Engine

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the Fire Protection System Diesel Engine

Functional test be changed from every 18 months (+25%) to every 24

months (+25%).

Except for periodic testing, the diesel is in a standby state

and not subject to operational stress. Periodic testing imposes

limited wear as evidenced by the absence of major repairs during

past maintenance. Extension of the operating cycle for several

months is expected to have virtually no impact upon diesel

operability. Monthly testing would detect any degradation. Thus it

is concluded that there would be no significant increase in the

probability or consequences of an accident as a result of an

extended interval between surveillances.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Extension of the plant operating cycle will, for the most part,

only extend the time spent by the pumps in standby capacity. The

potential for system deterioration is minimal under these

circumstances. Any deterioration that does occur will be slow acting

with respect to time. Significant deterioration in performance would

be detected by the monthly pump operating test. Thus, an extended

operating cycle is not expected to create the possibility of a new

or different kind of accident from any previously analyzed.

3. There has been no reduction in the margin of safety.

Extension of the operating cycle by several months only serves

to extend the period of time when the pumps are in standby status.

Any deterioration under these circumstances will be slow acting and

significant deterioration would be detected by the monthly operating

test. Therefore, it is concluded that an extended interval between

surveillances will involve no significant reduction in the margin of

safety.

(14) Electrical Tunnel, Diesel Generator Building, and

Containment Fan Cooler Fire Protection Spray Systems; (A) System

Functional Test and (B) Spray Header Visual Inspection

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

(a) It is proposed that the functional test surveillance

interval for the Electrical Tunnel, Diesel Generator Building and

Containment Fan Cooler Fire Protection Spray Systems be changed from

every 18 months (+25%) to every 24 months (25%).

(b) It is proposed that the Spray Header visual inspection

interval be revised from every 18 months (25%) to 24 months (+25%).

Extension of the surveillance interval for Electrical Tunnel and

Diesel Generator Building Fire Protection System functional tests

will have virtually no impact upon the operability of these systems.

These systems are accessible during normal operation and other

sections of the Technical Specifications (4.14.A.1.g.(i) and

4.14.B.1.a(i)) require that the system valve tests be conducted on

an annual (12 month) basis. These annual tests would reveal any

system deterioration prior to the conclusion of the proposed

extended surveillance interval.

For the Fan Cooler Fire Protection System as well as the Spray

Header itself, evaluation of surveillance data from the past five

refueling outages indicates minor discrepancies which would not have

impaired system operability.

It is therefore concluded that extension of the proposed

surveillance interval will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

For the Electrical Tunnel and Diesel Generator Building,

extension of the surveillance interval will have a negligible affect

as other portions of the Technical Specifications require the same

surveillance on an annual basis. For the spray header and the fan

cooler fire protection system, historical surveillance data

validates operability over an 18 month (+25%) interval which lends

confidence to conclude that operability will be maintained over a 24

month (+25%) interval. It is therefore concluded that the

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

previously evaluated has not been introduced.

3. There has been no reduction in the margin of safety.

Extension of the surveillance for two systems will have minimal

impact as the Technical Specifications impose more frequent testing

for system valves on an annual basis. For the Spray Header and Fan

Cooler Fire Protection System, as well as the fire protection system

for the Diesel Generator Building and Electrical Tunnel, it can be

stated that these systems are static existing mainly in a standby

capacity under which little deterioration would be expected. Past

surveillance data validates system reliability. It is therefore

concluded that increasing the time interval between inspections

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

(15) Penetration Fire Barriers

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the visual inspection frequency of the

penetration fire barriers listed in the Technical Specifications be

changed from every 18 months (+25%) to every 24 months (+25%).

The fire barrier penetration seals are static devices existing

in standby status. Normal environmental conditions exist during

normal plant operations. The only deterioration expected would be

that due to aging in a normal ambient which would be minimal to non-

existent. Evaluation of unacceptable seals detected during

surveillances indicates that initial seal installation was faulty

and aging was not the cause. Surveillances during four refueling

outages confirm this evaluation. Accordingly, it is not expected

that the proposed change in surveillance interval will involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Past surveillances indicate that time is not a predominate

failure mechanism. In the few unacceptable seals detected, the

initial installation procedure has been identified as the cause of

the problems. Since the seals are static devices which exist in a

standby condition and experience normal ambient conditions during

normal operation, this would be the expected conclusion. In

addition, the fire barriers are just one means of fire protection.

Other means of fire protection exist such as fire alarms, sprinklers

and heat detectors which provide defense in depth. Thus, it is

concluded that the proposed change in the surveillance interval will

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

from that previously evaluated.

3. There has been no reduction in the margin of safety.

Aging has not been identified as a principle contributor to seal

failures. In addition, there exists additional means of fire

protection which provides defense in depth. Therefore, the proposed

change in surveillance intervals is not expected to involve a

significant reduction in the margin of safety.

(16) Smoke Detectors/Electrical Penetration Area Inside

Containment

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the surveillance interval for the smoke

detectors located in the electrical penetration area inside

containment be revised from every 18 months (+25%) to every 24

months (+25%).

Based on data taken from six surveillances from 1984 through and

including 1993, these devices have proven to be highly reliable. No

test failures were observed during this period. Based on the

guidance contained in Generic Letter 91-04, this demonstration of

reliable performance provides an adequate basis to conclude that the

proposed extension in the surveillance interval will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Only 3 of the 5 detectors are required during normal operation.

Past surveillance data from six refueling outages indicate that it

is reasonable to expect all 5 detectors will remain operable over

the extended operating cycle which provides margin. It is therefore

concluded that the possibility of a new or different kind of

accident from any accident previously evaluated has not been

created.

3. There has been no reduction in the margin of safety.

The proven reliability of these devices indicates that a

significant reduction in the margin of safety would not be involved

in extending the operating cycle to 24 months (+25%).

(17) Functional Testing of Containment Sump Pumps

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the functional test of the Containment Sump

Pump be changed from every 18 months (+25%) to every 24 months

(+25%).

No credit is taken within the FSAR for the Containment Sump

Pumps as a means of mitigating the consequences of an accident.

During normal operation the pumps serve as a means of quantifying

leakage inside Containment and therefore serve a safety function in

terms of accident prevention. However, in this capacity they are

only one of several systems which are capable of serving this

function and their failure would not result in a loss of this

capability.

In addition, evaluation of surveillance data back to 1986

indicates, with one exception, that the devices are very reliable.

In one instance, the pumps did not actuate or cause operation within

the setpoint tolerance but did operate as required. This was

determined not to be a time dependent event.

It is therefore concluded that extending the interval between

refueling surveillances will not result in a significant increase in

the probability or consequences of an accident.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Past surveillances indicate that time is not a predominate

failure mechanism. Also, there exists a Technical Specification

requirement to perform almost the same surveillance on a monthly

basis in addition to every refueling outage. This monthly test

diminishes any potential risk in extending the operating cycle. It

is therefore concluded that the possibility of a new or different

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

3. There has been no reduction in the margin of safety.

Past surveillance data indicates that pump operation is

reliable. In addition, there are alternate means of providing the

safety function fulfilled by these pumps. Also, a monthly test is

required which would detect any malfunction prior to the end of an

extended operating cycle. It is therefore concluded that extending

the operating cycle by several months will not result in a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: Michael J. Case, Acting

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

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

Date of amendment request: November 2, 1994

Description of amendment request: The proposed amendments will

upgrade existing TS 3/4.6.2.1 and TS 3/4.6.2.3 by adapting the combined

specification for Containment Spray and Cooling Systems, contained in

the Standard Technical Specifications for Combustion Engineering

Plants, to the St. Lucie units. The changes account for plant-specific

differences and include all related requirements of NUREG-1432, Rev. O,

specification 3.6.6A. Accordingly, the proposal is consistent with the

Commission's Final Policy Statement on Technical Specifications

Improvements (58 FR 39132).

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:

Pursuant to 10 CFR 50.92, a determination may be made that a

proposed license amendment involves no significant hazards

consideration if 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. Each standard is discussed as

follows:

(1) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment will upgrade the existing Limiting

Conditions for Operation (LCOs) associated with the Containment

Cooling and Spray Systems to be consistent with NUREG-1432, Standard

Technical Specifications for Combustion Engineering Plants. The

Containment Cooling and Spray Systems are not initiators of

accidents previously evaluated, but are included as part of the

success paths associated with mitigating various accidents and

transients. The redundancy afforded by Containment Cooling and Spray

Systems in conjunction with the requirements of the proposed LCO

assures that the safety function of these systems can be

accomplished considering single failure criteria. Neither the design

nor the safety function of the Containment Cooling and Spray Systems

have been altered, and the proposed amendment does not change the

applicable plant safety analyses. Therefore, operation of the

facility in accordance with the proposed amendment will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the

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

administrative in nature in that they do not involve the addition of

new equipment or the modification of existing equipment, nor do they

otherwise alter the design of St. Lucie Unit 1 & 2 systems.

Therefore, operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The safety function of the Containment Cooling System is to

provide containment heat removal during normal operation and

accident conditions. The safety function of the Containment Spray

System is to provide containment heat and iodine removal during

accident conditions. The proposed amendment, in conjunction with the

redundancy afforded by the Containment Cooling and Spray system

design, assures that these safety functions can be accomplished

considering single-failure criteria. The bases for required actions

and the action completion times specified for inoperable Containment

Cooling and Spray trains are consistent with the corresponding

specifications in NUREG-1432. The safety analyses for applicable

accidents and transients remain unchanged from those previously

evaluated and reported in the Updated Final Safety Analysis Report.

Therefore, operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

Based on the above discussion and the supporting Evaluation of

Technical Specification changes, FPL has determined that the

proposed license amendment involves no significant hazards

consideration.

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: Indian River Junior College

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

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Mohan Thadani, Acting

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: November 7, 1994

Description of amendment request: The proposed amendment would

change the number of diesel generators (emergency power supply)

required to be operable during Mode 5 with the loops filled and Mode 6

with greater than or equal to 23 feet of water above the reactor vessel

flange. In addition, changes to certain system specifications that are

affected by the changes for the emergency power supply were also

proposed.

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 accidents previously

evaluated.

The equipment which is affected by the technical specification

changes proposed here are not precursors to any accident postulated

to occur in Modes 5 and 6. Therefore, the probability of an accident

is not increased. A design review has demonstrated the ability of

the required systems to perform their accident mitigation functions

for the postulated accidents during Mode 5 and 6 operation.

Therefore, it is concluded that an increase in the consequences of

the postulated accidents will not result from the proposed Technical

Specifications.

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

or different kind of accident from any previously evaluated.

The system design, function, and performance is not affected by

these specifications. No new equipment interactions are created.

Calculations and Failure Modes and Effects Analyses (FMEA) have been

conducted for selected mechanical systems and show there are no

failures which would cause situations where applicable accidents

would not be mitigated or which would cause new accidents. On this

basis, the proposed 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 the margin of safety.

The electrical power system specifications support the equipment

required to be operable, commensurate with the current level of

safety, including the equipment requiring a diesel backed power

source. The design review results demonstrate that operation in

Modes 5 and 6, in accordance with the proposed Technical

Specification changes, is acceptable from an accident mitigation

standpoint. The basic Modes 5 and 6 plant system functions are not

changed. On this basis, the proposed change does not involve a

significant reduction in the margin of safety.

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

this review, it appears that the 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, Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: William D. Beckner

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: November 7, 1994

Description of amendment request: The proposed amendment would

permit both containment personnel airlock doors to be open while moving

fuel during refueling operations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change to Technical Specification 3.9.4,

Containment Building Penetrations, would allow the containment

personnel airlock to be open during fuel movement and core

alterations. The containment personnel airlock is closed during fuel

movement and core alterations to prevent the escape of radioactive

material in the event of a fuel handling accident. The containment

personnel airlock is not an initiator to any accident. Whether the

containment personnel airlock doors are open or closed during fuel

movement and core alterations has no affect on the probability of

any accident previously evaluated.

The proposed change does alter assumptions previously made in

evaluating the radiological consequences of the fuel handling

accident inside the reactor containment building. The proposed

change allows for the containment personnel airlock to be open

during refueling. The radiological consequences described in this

change are bounded by those given in the South Texas Project Safety

Evaluation Report and General Design Criteria 19. All doses for the

proposed change are less than the acceptance criteria, therefore,

there is no significant increase in the consequences of an accident

previously analyzed.

The proposed change will significantly reduce the dose to

workers in the containment in the event of a fueling handling

accident by accelerating the containment evacuation process. The

proposed change will also significantly decrease the wear on the

containment personnel airlock doors and, consequently, increase the

reliability of the containment personnel airlock doors in the event

of an accident.

Since the probability of a fuel handling accident is unaffected

by the airlock door positions, and the increased doses do not exceed

acceptance limits, the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

The proposed change affects a previously evaluated accident,

e.g., a fuel handling accident inside containment. The existing

accident has been modified to account for the containment personnel

airlock doors being opened at the time of the accident. It does not

represent a significant change in the configuration or operation of

the plant and, therefore, does not create the possibility of a new

or different type of accident from any accident previously

evaluated.

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

margin of safety?

The margin of safety is reduced when the offsite and control

room doses exceed the acceptance criteria in the STP SER. As

previously discussed in the response to question 1, the offsite and

control room doses are below the acceptance criteria. Therefore,

this proposed change does not significantly reduce the 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, Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: William D. Beckner

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: November 8, 1994

Description of amendment request: The proposed amendment would

require only one of the two battery chargers associated with each Class

1E 125 VDC Channel I and Channel IV to be operable.

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 accidents previously

evaluated.

A single charger is able to maintain the operability of Channel

I or Channel IV at the design loading with a single failure

condition. The proposed change does not alter equipment or

assumptions made in previously evaluated accidents. The consequences

of previously evaluated accidents are not increased. On this basis,

the proposed change does not involve a significant increase in the

probability or consequences of accidents previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed change involves only the operability requirement

for the second battery charger in Channel I and Channel IV. The

failure modes and operating modes would then be identical for all

four STPEGS Class 1E DC channels. Failure modes and effects analyses

already performed for DC Channels II and III would thus become

applicable to Channels I and IV also. The change proposed by this

Technical Specification revision is bounded by the failure modes and

effects analysis provided as Table 8.3-8 of the STPEGS UFSAR

[Updated Final Safety Analysis Report]. On this basis, the proposed

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 the margin of safety.

The proposed change involves only the operability requirement

for the second battery charger in Channel I and Channel IV. The

number and capacity of DC channels required is not affected by the

proposed change. The electrical loads supported by these DC channels

are not changed and the duration of their function is not impacted.

On this basis, the proposed change does not involve a significant

reduction in the margin of safety.

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

this review, it appears that the 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, Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, NW, Washington, DC 20036

NRC Project Director: William D. Beckner

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: November 8, 1994

Description of amendment request: The proposed amendment would

permit the substitution of an extended range neutron flux monitor for

one of the source range neutron flux monitors during refueling

operations.

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 as previously

evaluated.

During refueling operations, the Source Range channels are used

only for monitoring changes in core reactivity, and does not provide

inputs for automatically actuated equipment. The same function could

be performed by an Extended Range channel. The combination of the

present Channel Check and the proposed Channel Calibration are

sufficient to ensure that the detectors are capable of monitoring

core reactivity changes. By providing the intended redundant core

reactivity monitoring, neither the possibility or consequences of an

accident previously evaluated are increased.

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

or different kind of accident from any previously evaluated.

During refueling operations, the Source Range Monitors are used

simply as monitoring instrumentation. Extended Range Monitors are

capable of performing this function. The combination of the present

Channel Check and the proposed Channel Calibration are sufficient to

ensure that the detectors are capable of monitoring core reactivity

changes.

The proposed change would require revision of STP refueling

procedures. However, the physical movement of fuel assemblies is

within the scope of current refueling procedures. No new mechanism

for fuel misloading or damage or boron dilution would be created by

the change. Therefore, the proposed 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 change provides core reactivity monitoring

comparable to that provided by the use of the Source Range channels.

The Extended Range channel is capable of detecting core reactivity

changes and provides the intended redundancy. The combination of the

present Channel Check and the proposed Channel Calibration are

sufficient to ensure that the detectors are capable of monitoring

core reactivity changes. No margin of safety is compromised by this

change.

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, Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: William D. Beckner

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine

YankeeAtomic Power Station, Lincoln County, Maine

Date of amendment request: October 24, 1994

Description of amendment request: The proposed amendment would

modify Technical Specifications Table 4.1-3 surveillance requirements

for new emergency feedwater flow instrumentation. Specifically, the

currently installed analog feedwater flow transmitters would be

replaced by new, digital-type flow transmitters. The new digital flow

emergency feedwater flow transmitters are continuously self-checking

and have a recommended calibration interval of 9 years. The licensee

proposes to verify flow whenever the system operates and send one

transmitter back to the manufacturer for recalibration every 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. The NRC staff has reviewed the licensee's analysis

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

presented below:

1. The proposed change does not involve a significant increase in

the

probability or consequences of an accident previously evaluated.

Performance of Technical Specifications Table 4.1-3 (items 10 a and

b) ensures the emergency feedwater flow transmitters are operable when

required. The proposed change will continue to ensure operabililty and

therefore will not increase 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.

Emergency feedwater flow transmitter operability verification is

maintained, with no change to the system's configuration. Thus, there

is no unique operating condition that could adversely affect system

functional performance.

3. The proposed change does not involve a significant reduction in

a margin of safety.

There is no change to any Final Safety Analysis Report Chapter 14

(Safety Analysis) event. There is no change to the demonstration of

component operability; thus, the proposed change does not involve a

significant reduction in a margin of safety.

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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine 04578

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, Maine 04011

NRC Project Director: Walter R. Butler

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

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

Date of amendment request: October 7, 1994

Description of amendment request: The proposed amendment would

remove from the Technical Specifications certain audit responsibilities

of the Nuclear Safety Audit Review Committee and certain review

responsibilities of the Station Operation Review Committee relating to

the Emergency Plan and Security Plan and their implementing procedures.

The proposed changes are consistent with the guidance of Generic Letter

93-07.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

A. The changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed changes do not affect the manner by

which the facility is operated and do not change any facility design

feature or equipment. Since there is no change to the facility or

operating procedures, there is no affect upon the probability or

consequences of any accident previously analyzed.

B. The changes do not create the possibility of a new or different

kindof accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because they do not affect the manner by which the

facility is operated. The proposed changes merely affect audit and

review responsibilities and their deletion or relocation to other

controlled documents does not introduce new or different accident

scenarios.

C. The changes do not involve a significant reduction in a margin

of safety (10 CFR 50.92(c)(3)) because the proposed changes do not

affect the manner by which the facility is operated or involve

equipment or features which affect the operational characteristics of

the facility.

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: Exeter Public Library, 47

Front Street, Exeter, NH 03833.

Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One

International Place, Boston, MA 02110-2624.

NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,

Millstone Nuclear Power Station, Unit 1, New London County,

Connecticut

Date of amendment request: October 14, 1994

Description of amendment request: The proposed change clarifies the

low pressure coolant injection (LPCI) requirements as required by

Technical Specification 4.5.A.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed change in accordance with 10 CFR

50.92 and concluded that the change does not involve a significant

hazards consideration (SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not compromised. The

proposed change does not involve an SHC because the change would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The LPCI flow surveillance requirement to demonstrate that three

pumps can deliver 15,000 gpm does not relate to any previously

analyzed accident. There are no accident scenarios which rely upon

three pumps or 15,000 gpm. The existing scenarios are more limiting

in that they rely on, at the most, two LPCI pumps. The actual

testing of the pumps in accordance with the [inservice testing] IST

program and Technical Specification 4.13 will not change. The

testing of the pumps currently performed demonstrates that LPCI will

function to mitigate the postulated accidents. Therefore, the

elimination of the requirement to demonstrate that three pumps can

deliver 15,000 gpm will not involve an increase in the probability

or consequence of any previously evaluated accident.

The elimination of a requirement to test the LPCI header

instrumentation can not result in an increase to the probability or

consequence of an accident, since no such instrumentation exists, or

is required.

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

from any previously analyzed.

The proposed change will remove a requirement to perform a

mathematical evaluation that provides no safety benefit. There is no

change in the test methodology currently performed. All four LPCI

pumps are tested. Deleting the requirement to verify that three LPCI

pumps can produce 15,000 gpm flow does not create the possibility of

a new or different type of accident.

Deleting the requirement to test the LPCI spray header

instrumentation can not create the possibility of a new or different

kind of accident, since there is no LPCI header instrumentation.

This change corrects an error which was introduced by an earlier

License Amendment.

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

This change to the LPCI testing requirements does not change any

of the actual testing, or individual component requirements which

exist for the LPCI system. The change to remove the three pump,

15,000 gpm flow requirement eliminates the need to calculate a value

which provides no relevant information in ascertaining the ability

of the LPCI system to perform its required safety function. The

existing testing ensures performance of the LPCI subsystem in

accordance with the accident analysis requirements. The intent of

the Technical Specification Surveillance Requirement remains

unchanged. Elimination of the requirement to test the LPCI header

instrumentation corrects an error introduced in an earlier License

Amendment. No LPCI header instrumentation exists, therefore, no

credit was taken for such instrumentation in determining the margin

of safety.

This change can not involve a significant reduction in the

margin of safety since there are no changes to the surveillance

requirements for any of the individual components of the LPCI

system.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee

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

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota

Date of amendment requests: October 17, 1994, as supplemented

October 27, 1994

Description of amendment requests: The proposed amendments would

change the submittal frequency of the Radioactive Effluent Release

Report from semiannual to annual in accordance with 10 CFR 50.36a.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated

The proposed license amendments are requested to implement a

revision to 10 CFR 50.36a. The requested amendment[s] does not alter

any administrative controls over radioactive effluents, nor do they

affect any accident evaluations. Also, the requested amendments do

not involve any physical alterations to the plant with respect to

radioactive effluents. The proposed changes would only affect the

reporting requirements concerning routine data for radioactive

effluents.

Therefore, the probability or consequences of an accident

previously evaluated are not affected by any of the proposed

amendments.

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

new or different kind of accident from any accident previously

analyzed

The proposed license amendments are requested to implement a

revision to 10C FR 50.36a. The requested amendment[s] does not alter

any administrative controls over radioactive effluents, nor do they

involve any physical alterations to the plant with respect to

radioactive effluents. Also, the requested amendments do not change

the method by which any safety-related system performs its function.

The proposed changes would only affect the reporting requirements

concerning routine data for radioactive effluents.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be

created.

3. The proposed amendment will not involve a significant

reduction in the margin of safety

The proposed license amendments are requested to implement a

revision to 10 CFR 50.36a. The requested amendment[s] does not alter

any administrative controls over radioactive effluents, nor do they

involve any physical alterations to the plant with respect to

radioactive effluents. The proposed changes would only affect the

reporting requirements concerning routine data for radioactive

effluents. The operation of systems and equipment remains unchanged.

Therefore, a significant reduction in the margin of safety would

not be involved.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: October 25, 1994

Description of amendment request: The amendment would add to the

Susquehanna Units 1 and 2 Technical Specifications, isolation signals

to Table 3.6.3-1 for the containment isolation valves on the sample

lines for the containment radiation monitoring (CRM) and wetwell sample

lines. This change is based on the licensee's design change for

installation of a new CRM and wetwell sample 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. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The addition of the new CRM and Wetwell Sample System does not

affect any of the postulated initiating events identified in Chapter

6 and 15 of the FSAR, the Design Assessment Report, the current

Reload Analysis or the NRC Safety Evaluation Report (NUREG 0776).

The new CRM and Wetwell Sample System with separate containment

sample lines is isolated from the primary containment under accident

conditions. The power and control-power to the CRM from the Class 1E

Division I and Division II sources is through electrical isolation

schemes so that failure(s) in the CRM under accident conditions is

isolated from the Class 1E systems.

The addition of a new CRM and Wetwell Sample System with

separate sample lines and isolation valves does represent a change

in the probability of occurrence of a malfunction of equipment. The

addition of the auxiliary relay to the Division I and Division II

CAC System containment isolation logic does represent the source of

another potential malfunction in the logic due to the additional

relay in the circuit. However, the increase in probability due to

the additional relay is considered to be so small or insignificant

that the change is within the error bounds associated with the

original design calculations and does not constitute a significant

increase in probability of the overall system malfunction.

Thus, the addition of a new CRM and Wetwell Sample System does

not significantly increase the probability of occurrence or the

consequences of an accident or malfunction of equipment important to

safety, as previously evaluated in the SAR.

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

from any accident previously evaluated.

Chapter 6 and 15 of the FSAR, the Design Assessment Report, the

current Reload Analysis and NUREG-0776 were reviewed to determine if

the proposed action had the potential of creating a postulated

initiating event which was not within the spectrum of events which

transient or anticipated operational occurrences and accident

conditions were analyzed. The review did not identify a postulated

initiating event which would create the possibility for an accident

of a different type.

A random single failure in the CRM A or CRM B does not create a

malfunction of a different type. A random single failure in the

existing containment isolation circuitry, the new isolation valve

control circuitry or the new valve position indication circuitry for

the new containment isolation valves does not create a malfunction

of a different type. The consequences of random single failure of

the CRM or the CRM and Wetwell containment isolation valve isolation

signal, control and indication circuitry is the same as the existing

consequences.

Thus, the addition of a new CRM and Wetwell Sample System does

not create a possibility for an accident or malfunction of a new or

different type.

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

The operability of the primary containment isolation valves for

the sample lines to the new CRMs and Wetwell Sample Rack is governed

by Technical Specification Section 3/4.6.3 entitled ``Containment

Systems, Primary Containment Isolation Valves'' with Table 3.6.3-1

establishing the maximum isolation time. The bases for operability

of the primary containment isolation valves is to ensure that the

containment atmosphere is isolated from the outside environment in

the event of a release of radioactive material to the containment

atmosphere or pressurization of the containment. This is consistent

with GDC 54 through 57 of 10 CFR 50, Appendix A. The bases for the

containment isolation within the time limits specified in Table

3.6.3-1 is for those isolation valves designed to close

automatically to ensure that the release of radioactive material to

the environment is consistent with the assumptions used in the

analyses for a LOCA. The new CRM and Wetwell Sample Rack sample line

isolation valves are solenoid valves which close immediately on an

accident signal. The proposed action does not affect the operability

requirements of Section 3/4.6.3. The margin of safety as defined in

the Technical Specification for the containment isolation valves is

not affected.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: November 11, 1994

Brief description of amendments: The proposed amendment would

modify the technical specifications (TS) by deleting accelerated

testing and special reporting requirements for CPSES Units 1 and 2

emergency diesel generators. These changes are based on Generic Letter

94-01, ``Removal of Accelerated Testing and Special Reporting

Requirements for Emergency Diesel Generators,'' dated May 31, 1994.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

Deletion of the requirement for special reporting of EDG

failures has no relation to probability or consequences of

accidents. Therefore, deletion of the requirement for special

reporting of EDG failures does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

There are no initiating events in accidents previously evaluated

that involve testing of EDGs. Therefore, deletion of accelerated

testing of EDGs does not involve a significant increase in the

probability of an accident previously evaluated.

A reduction in the number of test starts decreases EDG component

stress and wear and decreases unavailability time for maintenance

and pre and post run checks. The resulting change in EDG reliability

and availability is an improvement toward ensuring the EDGs are

capable of fulfilling their functional requirement to provide

electric power for safe shutdown of the plant during loss of offsite

power. Furthermore, implementation of the maintenance rule

provisions for performance monitoring and root cause analysis for

failures as a basis for establishing corrective actions establish an

alternate reliability basis that is at least equivalent to that

established by accelerated testing. Therefore, deletion of

accelerated testing of EDGs does not involve a significant increase

in the consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

Deletion of the requirement for special reporting of EDG

failures introduces no new failure modes for the EDGs or other plant

systems and therefore has no relation to creation of accidents.

Therefore, deletion of the requirement for special reporting of EDG

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

of accident from any accident previously evaluated.

The frequency at which EDG testing occurs does not affect the

potential failure modes of the EDGs, which have already been

assessed in the CPSES design. Therefore, deletion of accelerated

testing of EDGs does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

margin of safety?

Acceptance limits and failure values are not affected by the

requirement for special reporting of EDG failures. Therefore,

deletion of the requirement for special reporting of EDG failures

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

The margin of safety impact associated with accelerated testing

relates to EDG reliability and availability. A reduction in the

number of test starts decreases EDG component stress and wear and

decreases unavailability time for maintenance and pre and post run

checks. The resulting change in EDG reliability and availability is

an improvement toward ensuring the EDGs are capable of fulfilling

their functional requirement to provide electric power for safe

shutdown of the plant during loss of offsite power. Furthermore,

implementation of the maintenance rule provisions for performance

monitoring and root cause analysis for failures as a basis for

establishing corrective actions establish an alternate reliability

basis that is at least equivalent to that established by accelerated

testing. Therefore, deletion of accelerated testing of EDGs 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: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036

NRC Project Director: William D. Beckner

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: November 11, 1994

Brief description of amendments: The proposed amendment would

provide for cycle-specific allowances to account for increases in the

Heat Flux Hot Channel Factor between monthly surveillances.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed changes provide for the use of cycle-specific

allowances to account for F2Qc(z) increases

between surveillances. No hardware or setpoint changes are involved;

therefore, the changes have no impact on the probability of

occurrence of any accident previously analyzed.

The proposed changes ensure that F2Qc(z)

remains within its limit. Thus, the changes do not increase the

consequences of any accident previously analyzed.

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

different kind of accident from any accident previously evaluated?

The proposed changes provide for the use of a cycle-specific

allowances to account for F2Qc(z) increases

between surveillances. The proposed changes do not involve any

hardware or setpoint changes. Therefore the changes do not create

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

accident previously analyzed.

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

margin of safety?

The proposed changes do not affect the failure values of any

system or any event acceptance criteria. Higher cycle-specific

allowances ensure that remains below its limit between surveillances

and within the bounds considered in the safety analyses. Therefore,

the proposed changes do 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: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036

NRC Project Director: William D. Beckner

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

Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: November 8, 1994

Description of amendment request: The proposed amendment would

revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications

to allow application of a voltage-based repair limit for the steam

generator (SG) tube support plate (TSP) intersections experiencing

outside diameter stress corrosion cracking (ODSCC).

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:

This proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Testing of model boiler specimens for free span tubing (no TSP

restraint) at room temperature conditions show burst pressures in

excess of 5,000 psig for indications of ODSCC with voltage

measurements as high as 19 volts. Burst testing performed on five

intersections pulled from the Kewaunee SGs with up to a 2 volt

indication showed measured tube burst in the range of 9,537 to 9,756

psig. Burst testing performed on pulled tubes from other plants with

up to 7.5 volt indications show burst pressures in excess of 6,300

psi at room temperatures. Correcting for the effects of temperature

on material properties and the minimum strength levels, tube burst

capability significantly exceeds the safety factor requirements of

RG 1.121.

Tube burst criteria are inherently satisfied during normal

operating conditions due to the presence of the TSPS. Test data

indicates that tube burst cannot occur within the TSP, even for

tubes with through wall EDM notches 0.75 inch long, when the notch

is adjacent to the TSP. Since tube burst is precluded during normal

operating conditions, the criterion that must be satisfied to

demonstrate adequate tube integrity is a safety margin of 1.43 times

MSLB pressure differential. From Figure 3-2 of EPRI report TR-

100407, the BOC structural limit for 7/8 inch diameter tubing is 9.6

volts. Applying an allowance of 20% for NDE uncertainty and 50% for

crack growth rate over an operating cycle results in a voltage

repair limit of 5.6 volts. The proposed repair limit of 2 volts is

very conservative when compared to the 5.6 volts taking into account

the low average growth rates experienced at Kewaunee and the high

tube burst pressures.

Relative to the expected leakage during accident condition

loadings, a plant specific calculation was performed to determine

the maximum primary-to-secondary leakage during a postulated MSLB

event. The evaluation considered both pre-accident and accident

initiated iodine spikes. The results of the evaluation show that the

accident spike yielded the limiting leak rate. This case was based

on a 30 rem thyroid dose at the site boundary and initial primary

and secondary coolant activity levels of 1.0 uCi/gm and 0.1 uCi/gm

dose equivalent iodine-131, respectively. A leak rate of 34.0 gpm

was determined to be the upper limit for allowable primary to

secondary leakage in the SG in the faulted loop. The SG in the

intact loop was assumed to leak at a rate of 0.1 gpm (150 gpd).

Application of the voltage-based repair limit will be

supplemented with a projected EOC MSLB leakage calculation and

conditional burst probability assessment. The methodology for

performing these calculations will be consistent with that discussed

in the draft GL until final guidance is published. Should the

projected MSLB leakage be exceeded indications will be repaired or

removed from service until the projected leakage is less than or

equal to 34.0 gpm.

Application of the voltage-based repair limit will not adversely

affect SG tube integrity. Therefore, the proposed amendment will not

increase the probability or consequences of an accident previously

evaluated.

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

from any previously evaluated.

Implementation of the proposed voltage-based repair limit will

not reduce the overall safety or functional requirements of the SG

tube bundles. The tube burst criteria will be satisfied during

normal operating conditions by the presence of the TSPs. The RG

1.121 criteria that must be satisfied during accident loading

conditions is 1.43 times MSLB differential pressure. Conservatively,

the existing data base of burst testing shows that the tube burst

margins can be satisfied with bobbin coil signal amplitudes of about

8.82 volts or less regardless of the depth of tube wall penetration.

The proposed repair criteria will be supplemented with a reduced

operating leakage requirement of 150 gpd average through either SG

to preclude the potential for excessive leakage during operating

conditions. The 150 gpd restriction will provide for timely 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. The operating leakage limit is

based on leak-before break considerations, critical crack length and

predicted leakage.

The SG tube integrity will continue to be maintained through

inservice inspections and primary-to-secondary leakage monitoring.

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

new or different kind of accident.

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

Application of the voltage-based repair criteria has been

demonstrated to maintain tube integrity commensurate with the RG

1.121 criteria. RG 1.121 describes a method acceptable to the staff

for meeting GDCs 2, 14, 15, 31 and 32. This is accomplished by

determining the limiting degradation of SG tubing as established by

inservice inspection, beyond which tubes should be removed from

service. Upon implementation of the repair criteria, even under the

worst case conditions, the occurrence of ODSCC at the TSPs is not

expected to lead to a SG tube rupture event during normal or faulted

conditions. The most limiting event would be a potential increase in

leakage during a MSLB event. Excessive leakage during a MSLB is

precluded by verifying that the expected EOC crack distribution of

ODSCC indications at TSP locations would result in an acceptably low

primary-to-secondary leakage. Therefore, the radiological

consequences from tubes remaining in service is a small fraction of

the 10 CFR 100 limits.

The combined effects of a LOCA plus SSE on the SGs were assessed

as required by GDC 2. This issue was addressed for the Kewaunee SGs

through the application of leak-before-break (LBB) principles to the

primary loop piping. Based on the results of this analysis, it is

concluded that the LBB is applicable to the Kewaunee primary loops

and, thus, the probability of breaks in the primary loop piping is

sufficiently low that they need not be considered in the structural

design basis of the plant. Excluding breaks in the primary loops,

the LOCA loads from the large branch lines were also assessed and

found to be of insufficient magnitude to result in SG tube collapse.

Based on these analysis results, no tubes are expected to collapse

or deform to the degree that secondary-to-primary in-leakage would

be increased over currently expected levels. On this basis no tubes

need to be excluded from the voltage-based repair criteria for

reasons of deformation resulting from combined LOCA and SSE

loadings.

Addressing the RG 1.83 considerations, implementation of the

voltage-based repair criteria will include a 100% bobbin coil probe

inspection of all TSP intersections with known ODSCC down to the

lowest cold leg TSP identified. This will be supplemented by a

reduced operating leakage limit, enhanced eddy current data analysis

guidelines, MRPC inspection requirements and a projected EOC voltage

distribution. It is concluded that the proposed change will not

result in 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: University of Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: Leif J. Norrholm

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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

Project No. 2, Benton County, Washington

Date of application for amendment: August 8, 1994

Brief description of amendment request: The proposed amendment

would modify Technical Specification (TS) 4.0.5.a. to delete the

requirement to obtain prior written relief from the NRC for inservice

inspection (ISI) and inservice testing (IST) of components conducted

pursuant to 10 CFR 50.55a. This change would provide relief from the

ASME Code requirement in the interim between the submittal of a relief

request and the NRC's issuance of a safety evaluation regarding the

relief request. The change would allow the plant to operate in

accordance with a proposed relief request while the NRC staff completed

its review of the relief request. The licensee has also proposed to

modify TS 4.0.5.b. to add a definition for biennial or every-2-year

inspection and testing activities. The definition of biennial or every

2 years will be at least once per 731 days.Date of individual notice in

Federal Register: November 14, 1994 (59 FR 56558)

Expiration date of individual notice: December 14, 1994

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

SteamElectric Plant, Unit No. 2, Darlington County, South

CarolinaDate of application for amendment: June 29, 1994

Brief description of amendment: The amendment deletes the

requirement to perform alternate train testing to demonstrate that

other, similar, safety-related components are operable when components

are found, or made, inoperable in the safety injection, residual heat

removal, and containment spray systems. The surveillance requirements,

which the licensee refers to as accelerated testing requirements,

affect the safety injection (SI) pumps, residual heat removal (RHR)

pumps, containment spray (CS), SI, RHR and CS flow paths.

Date of issuance: November 21, 1994

Effective date: November 21, 1994

Amendment No. 153

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 3, 1994 (59 FR

39581) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 21, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College, Hartsville, South Carolina 29550

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: June 29, 1992, as supplemented

February 22, 1994

Brief description of amendment: The amendment revised TS Sections

3/4.3, ``Instrumentation,'' 3/4.4.2, ``Safety/Relief Valves,'' and

associated Bases to increase the surveillance test intervals and

allowable out-of-service times for specific safety-related

instrumentation.

Date of issuance: November 22, 1994

Effective date: November 22, 1994

Amendment No. 67

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

Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (59 FR

17605) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 22, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power

Station,Units 1 and 2, Rock Island County, IllinoisDate of

application for amendments: October 15, 1992

Brief description of amendments: The proposed amendments would

revise the Dresden and Quad Cities Technical Specification (TS) 3/4.4

to revise the sodium pentaborate solution concentrations for the

Standby Liquid Control System (SLCS) storage tanks based on net

positive suction head test results.

Date of issuance: November 16, 1994

Effective date: November 16, 1994

Amendment Nos.: 130, 124, 151, and 147

Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 9, 1992 (57 FR

58245) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 16, 1994. No

significant hazards consideration comments received: No

Local Public Document Room location: For Dresden, The Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450; For Quad Cities,

The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: October 29, 1993, as

supplemented on March 28, 1994, and November 8, 1994.

Brief description of amendment: The amendment revises the

surveillance intervals for the Volume Control Tank Level Instrument,

the Containment High Range Radiation Monitors, the Safety Injection

System Electrical Loading, the Safety Injection System, and the Reactor

Coolant System Subcooling Margin Monitors to accommodate a 24-month

fuel cycle. These revisions are being made in accordance wih the

guidance provided by Generic Letter 91-04, ``Changes in Technical

Specification Surveillance Intervals to Accommodate a 24-Month Fuel

Cycle.''

Date of issuance: November 16, 1994

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

30 days.

Amendment No.: 178

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 20, 1994 (59 FR

37067) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 16, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of application for amendments: August 25, 1994

Brief description of amendments: The amendments revise the testing

interval for auxiliary feedwater (AFW) system pumps from monthly to

quarterly on a staggered test basis. The amendments are consistent with

the guidance in NUREG-1366, ``Improvements to Technical Specifications

Surveillance Requirements'' and Generic Letter 93-05, ``Line-Item

Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation.'' In addition, a note

is incorporated from NUREG-1431, ``Revised Standard Technical

Specifications, Westinghouse Plants'' into the TS clarifying that the

turbine-driven AFW pump cannot be tested until the required pressure

exists in the secondary side of the steam generator.

Date of issuance: November 9, 1994

Effective date: November 9, 1994

Amendment Nos.: 151 and 133

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

revised the Technical Specifications.

Date of initial notice in Federal Register: September 28, 1994 (59

FR 49426) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 9, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Florida Power and Light Company, Docket No. 50-335, St. Lucie

Plant, Unit No. 1, St. Lucie County, Florida

Date of application for amendment: March 19, 1993, augmented August

18, 1994.

Brief description of amendment: This amendment allows a reduction

in Reactor Coolant System design flowrate from the current value of

370,000 gpm to 355,000 gpm in Technical Specifications Figure 2.1-1 and

Tables 2.2-1 and 3.2-1.

Date of issuance: November 25, 1994

Effective date: November 25, 1994

Amendment No.: 130

Facility Operating License No. DPR-67: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 28, 1993 (58 FR

25855) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 25, 1994. No

significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: April 29, 1994, as supplemented by

letter dated September 8, 1994.

Brief description of amendments: The amendments revise the

technical specifications to permit revision of the maximum allowable

power range neutron flux high setpoint when one or more main steam

safety valves are inoperable. In addition, new algorithm used to

calculate the revised setpoint values is incorporated into the Bases

for the technical specifications.

Date of issuance: November 22, 1994

Effective date: To be implemented within 30 days of issuance

Amendment Nos.: Unit 1 - Amendment No. 66; Unit 2 - Amendment No.

55

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

revised the Technical Specifications.

Date of initial notice in Federal Register: June 8, 1994 (59 FR

29628) The additional information contained in the supplemental letter

dated September 8, 1994, was clarifying in nature and, thus, within the

scope of the initial notice and did not affect the staff's proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated November 22, 1994.No significant hazards consideration

comments received: No

Local Public Document Room location: Wharton County Junior College,

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

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

Center,Linn County, Iowa

Date of applications for amendment: June 4, 1993, as supplemented

February 4, 1994, and May 6, 1994.

Brief description of amendment: The amendment revised the Technical

Specifications (TS) by changing the requirements of the TS Section 3.6,

``Primary Systems Boundary,'' adding definitions into Section 1.0,

``Definitions,'' and revising Bases Section 3/4.6. These changes

improved clarity and provided consistency of the TS with the Standard

TS (NUREG-1202). Typographical and administrative corrections were also

made in Section 3.6.

Date of issuance: November 17, 1994

Effective date: November 17, 1994, and to be implemented within 120

days

Amendment No.: 203

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

Technical Specifications.

Date of initial notice in Federal Register: July 21, 1993 (58 FR

39052). The May 6, 1994, application, repeated a TS change included in

the June 4, 1993, application, and proposed changes to the TS Bases.

The information in the February 14, 1994, supplement, did not change

theinitial no significant hazards determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated November 17, 1994. No significant hazards consideration comments

received: No.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

Linn County, Iowa

Date of application for amendment: July 12, 1994

Brief description of amendment: The proposed amendment changes the

requirement to perform the surveillance test for the channel functional

test Rod Block Monitor, Flow-biased Average Power Range Monitor and

Recirculation Flow instruments from within 24 hours prior to startup to

after the reactor is in the RUN mode, but prior to when each system is

assumed to function in the plant safety analysis.

Date of issuance: November 18, 1994

Effective date: November 18, 1994, to be implemented within 30 days

of issuance

Amendment No.: 204

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

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45025) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 18, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

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

South Carolina Electric & Gas Company, South Carolina Public

ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: October 29, 1993, as

supplemented March 11, 1994, May 18, 1994, September 20, 1994, and

October 20, 1994

Brief description of amendment: The proposed changes support the

installation of new steam generators at Summer Station. The changes

involve:

(1) alterations to the core operating limits

(2) changes to various reactor trip setpoints

(3) deletion of the negative flux rate trip

(4) removal of references to specific correlations used in the

departure from nucleate boiling (DNB) analyses

(5) changes to the steam/feedwater flow mismatch activation

specification

(6) changes to shutdown limits

(7) changes to instrument uncertainty allowances

(8) a change to the methodology for reactor coolant system (RCS)

flow determination

(9) modifications to DNB parameters

(10) a change to the engineered safety features actuation system

setpoints for steam generator water levels

(11) removal of the F* and L* criteria (12) addition of a

requirement for a first inservice inspection for the new steam

generators

Date of issuance: November 18, 1994

Effective date: November 18, 1994

Amendment No.: 119

Facility Operating License No. NPF-12. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 16, 1994 (59

FR 7968) The May 18, 1994, September 20, 1994, and October 20, 1994

submittals contained explanatory information and did not change finding

of nos significant hazards consideration as published in the FEDERAL

REGISTER. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 18, 1994.No significant

hazards consideration comments recei

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