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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice Applications and Amendments to Facility Operating

Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 23, 1993, through January 6, 1994.

The last biweekly notice was published on January 5, 1994 (59 FR 615).

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 P-223, Phillips Building, 7920 Norfolk Avenue,

Bethesda, 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 20555.

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

intervene is discussed below.

By February 18, 1994, 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 20555 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 20555, 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 20555, and at the local public document

room for the particular facility involved.

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

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

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: October 26, 1993

Description of amendment requests: The proposed amendment would

change the maximum nominal enrichment of the fuel allowed to be used in

the reactor core. Specifically, in Technical Specification 5.3.1,

``Fuel Assemblies,'' the fuel enrichment would change from ``a maximum

enrichment of 4.05 weight percent U-235'' to ``a maximum radially

averaged enrichment of 4.30 weight percent U-235 at any axial

location.''

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:

Standard 1 -- Involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The adequacy of a given core design must be demonstrated for each

core prior to core reloading. The fuel enrichment is only one of the

factors that must be considered in this determination. The fuel

enrichment does not directly impact the results of the plant safety

analysis.

The Unit 1, 2, and 3 fuel and storage areas have been analyzed

for a maximum radially averaged enrichment of any axial enrichment

zone with a fuel assembly of 4.30 w/o U-235. The criticality

analysis for Palo Verde's spent fuel pool is presented in Section

D.3, Criticality Safety Analysis. The results of these analyses

indicate that handling and storage of such fuel assemblies do not

involve an unreviewed safety question. The results of these analyses

are within the acceptance criteria defined in TS 5.6.1,

``Criticality.''

The applicable codes, standards and regulations of criticality

safety for spent fuel and new fuel storage include the following:

- General Design Criterion 62 - Prevention of Criticality in

Fuel Storage and Handling.

- NUREG -0800, USNRC Standard Review Plan, Section 9.1.2, Spent

Fuel Storage and Section 9.1.1, New Fuel Storage.

- ANSI/ANS-57.2-1983, ``Design Requirements for Light Water

Reactor Spent Fuel Storage Facilities at Nuclear Power Plants,''

Section 6.4.2

- ANSI/ANS-57.3-1983, ``Design Requirements for New Fuel Storage

Facilities at Light Water Reactor Plants,'' Section 6.2.4

- Qualification of Analytical Methodology Used In Spent Fuel

Storage Rack Analyses, CE Benchmark, 590962-PHD-004 Revision 0 dated

December 3, 1979.

These regulations and guides require that for spent fuel racks

the maximum calculated keff, including margin for uncertainty

in calculational method and mechanical tolerances, be less than or

equal to 0.95 with a 95% probability at a 95% confidence level.

In order to assure the true reactivity will always be less than

the calculated reactivity, the following conservative assumptions

were made for spent fuel rack criticality analysis:

- Pure, unborated water at 68 degrees Fahrenheit is used in all

calculations,

- An infinite array with no radial or axial leakage is modeled,

and

- Neutron absorption from spacer grids is neglected, i.e.,

replaced by water.

For the new fuel vault, a dual criteria applies in which the

maximum calculated keff, including uncertainties, is less than

or equal to 0.95 when flooded and less than or equal to 0.98 under

conditions of ``optimum moderation.''

Because the new fuel vault is normally dry, and low density

moderation of ``optimum moderation'' produces strong coupling

between assemblies, the following conservative assumptions are used:

- The storage array was enclosed on all six sides by a tight

fitting two foot concrete reflector,

- Unborated water is introduced uniformly throughout the storage

array and the space between fuel pins,

- Water density is varied uniformly from flooded to dry,

- Neutron absorption from spacer grids is neglected, i.e.,

replaced by water.

In the new fuel vault criticality analysis, leakage is

explicitly modeled, because the assumption of an infinite array with

no radial or axial leakage is unrealistic under conditions of low

density moderation. Leakage suppresses criticality at low moderator

density. Without 3-D modeling of the array, erroneously high values

of keff are calculated. Thus, the assumption on array leakage

is relaxed, but reflection from the walls, floor and ceiling is

included.

In addition to the above discussion of the new fuel vault

criticality analysis, the following conservative assumptions are

applied to both analyses:

- No credit is taken for the presence of burnable poison rods.

These rods displace fuel rod positions and are an integral part of

selected fuel assemblies.

- An upper bound for the fuel density was included in the

nominal configuration of the fuel densities.

- The upper statistical bound of the fuel assembly enrichment,

as based on the fuel fabrication specification, is included in the

statistical evaluation of uncertainties.

The criticality analysis of the Palo Verde new and spent fuel

racks shows that the maximum radially averaged fuel enrichment of

any axial enrichment zone within a fuel assembly which meets the

appropriate NRC limit with uncertainties is higher than 4.30 w/o U-

235.

Although a higher enrichment fuel cycle may result in fuel

burnup consisting of a slightly different mixture of nuclides and

inventory, the effect is insignificant because the isotopic mixture

and inventory of an irradiated assembly is relatively insensitive to

the fuel assembly's initial enrichment. Therefore, the doses from

postulated accidents are not significantly affected and continue to

be acceptable.

Standard 2 -- Create the possibility of a new or different kind

of accident from any accident previously analyzed.

Operation of Palo Verde with the proposed enrichment limit

change will not create any new or different kinds of accidents from

those previously evaluated.

The adequacy of a given core design shall be demonstrated for

each core prior to reloading. The fuel enrichment is only one of the

factors that must be considered in this determination.

Fuel handling and storage of fuel with radially averaged

enrichment of any axial enrichment zone within a fuel assembly of

4.30 w/o U-235 does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

Standard 3 -- Involve a significant reduction in a margin of

safety.

This amendment request will not involve a significant reduction

in a margin of safety.

The evaluation performed for each reload core assures that the

core design meets appropriate safety limits, including a

consideration of a significant reduction in the margin of safety.

See response provided in Standard 1 for information pertaining to

the demonstration of the adequacy of each core design.

Criticality analyses for fuel assemblies with a maximum radially

averaged enrichment (of any axial enrichment zone) of 4.30% U-235

for the Palo Verde Fuel Pool configurations presented in this

proposed Technical Specification amendment meet the criticality

acceptance criterion for Keff listed in Technical Specification

5.6.1.1. Technical Specification 5.6.1.1 states:

The spent fuel storage racks are designed and shall be

maintained with:

a. A Keff equivalent to less than or equal to 0.95 when

flooded with unborated water, which includes a conservative

allowance of 2.6% delta k/k for uncertainties as described in

Section 9.1 of the FSAR.

. . .

Based on the above evaluation, this proposed change does not

constitute a significant hazards consideration.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

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

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

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

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

STN 50-529, Palo Verde Nuclear Generating Station, Units 1 and 2,

Maricopa County, Arizona

Date of amendment requests: October 27, 1993

Description of amendment requests: The proposed amendment would

revise Technical Specification (TS) 6.9.1.8 to change the frequency for

submitting the Radioactive Effluent Release Report from semiannual to

annual, as allowed by the revised 10 CFR 50.36a requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Standard 1 -- Involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes are administrative in nature and do not involve

any change to the configuration or method of operation of any plant

equipment that is used to mitigate the consequences of an accident.

Also, the proposed changes do not alter the conditions or

assumptions in any of the FSAR accident analyses. Since the FSAR

accident analyses remain bounding, the radiological consequences

previously evaluated are not adversely affected by the proposed

changes. Therefore, it can be concluded that the proposed changes do

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Standard 2 -- Create the possibility of a new or different kind

of accident from any accident previously analyzed.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes are administrative in nature and do not involve

any change to the configuration or method of operation of any plant

equipment that is used to mitigate the consequences of an accident.

Accordingly, no new failure modes have been defined for any plant

system or component important to safety nor has any new limiting

failure been identified as a result of the proposed changes. Also,

there will be no change in the types or increase in the amount of

effluents released offsite. Therefore, it can be concluded that the

proposed changes do not create the possibility of a new or different

kind of accident from any accident previously evaluated.

Standard 3 -- Involve a significant reduction in a margin of

safety.

The proposed changes do not involve a significant reduction in a

margin of safety. The proposed changes are administrative in nature

and do not adversely impact the plant's ability to meet applicable

regulatory requirements related to liquid and gaseous effluents, and

solid waste releases. The proposed changes would also eliminate an

unnecessary burden of governmental regulation without reducing

protection for public health and safety. Therefore, it can be

concluded that the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

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

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

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

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

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

County, Maryland

Date of amendments request: December 8, 1993

Description of amendments request: As an active participant in the

industry-NRC improved Standard Technical Specifications (STS)

implementation effort, Baltimore Gas and Electric Company (BG&E)

volunteered to develop criteria for determining what requirements are

appropriate for inclusion in the Design Features section of the

Technical Specifications (TSs), and to submit a lead plant license

amendment applying those criteria. This proposed license amendment

justifies the adoption of the ``Design Features'' section of the

Combustion Engineering STS for the Calvert Cliffs Nuclear Power Plant

and then applies the developed criteria to the STS. The proposed

criteria for determining what requirements should be placed in the

Design Features section of the TSs are:

1. The amount, kind, and source of special nuclear material

required;

2. The place of the use of the special nuclear material; and

3. Those features of the facility such as materials of construction

and geometric arrangements, which, if altered or modified, would have

an immediate and significant effect on safety and are not covered in

the safety limits, limiting conditions for operation or surveillance

requirements of the Technical Specifications.

The Design Features section is Section 5.0 for the Calvert Cliffs

TS and Section 4.0 for the STS. Specifically, the following changes are

proposed to the Calvert Cliffs TSs and the STS for the Design Features

section.

Section 5.1 Site

Sections 5.1.1 and 5.1.2 contain maps of the site boundary and low

population zones, respectively. These maps also show the major

structures, effluent release points, meteorological tower location, and

the minimum exclusion area radius. The STS requires inclusion of the

site and exclusion area boundaries and the low population zone and

allows descriptions instead of maps. These maps or descriptions do not

meet any of the criteria. However, Criteria 2 requires that the place

of use of the special nuclear material be described. BG&E proposes to

delete the existing Sections 5.1.1 and 5.1.2 and to create Section 5.1,

entitled ``Site Location,'' and to include a text description of the

location of the site. The current information and maps in the sections

will be relocated to the Updated Final Safety Analysis Report (UFSAR).

This change is also proposed for Section 4.1 of the STS.

Section 5.2 Containment

This section does not meet Criteria 1, 2 or 3 in that modification

of the containment would not create an immediate and significant effect

on safety. Furthermore, containment integrity is covered in the

Limiting Conditions for Operation and Surveillance Requirements

sections of the TS. This Section is not included in the STS. BG&E

proposes to eliminate it from the Calvert Cliffs Design Features.

Section 5.3 Reactor Core

Calvert Cliffs TS Section 5.3 and corresponding STS Specification

4.2.1 and meets Criteria 1, e.g., state the amount, kind, and source of

special nuclear material. For consistency with the STS, BG&E proposes

to adopt the STS titles and wording.

Section 5.3.3 Control Element Assemblies

Calvert Cliffs Section 5.3.3, ``Control Element Assemblies,'' and

the corresponding STS Section 4.2.2 does not meet Criteria 1, 2 or 3.

The safety significant aspects of control rods, e.g., the reactivity

worth of control rods and their required insertion times, are included

in other portions of the TSs. Therefore, control rods do not fall under

the Criteria 3 as features not described in other sections of the TSs.

BG&E proposes that this section be eliminated from the Calvert Cliffs

Design Features and the STS.

Section 5.4 Reactor Coolant System

This section does not meet any of the criteria and is not included

in the STS. It does not meet Criteria 3 in that the requirements on

degradation, pressure, and temperature are contained in other portions

of the TS and changes in the total water and steam volume would not

have an immediate and significant impact on safety. BG&E proposes that

it be eliminated from the Calvert Cliffs Technical Specifications.

Section 5.5 Meteorological Tower Location

This section does not meet any of the criteria and is not included

in the STS. BG&E proposes to eliminate this section from the Calvert

Cliffs Design Features section.

Section 5.6 Fuel Storage

Calvert Cliffs Section 5.6.1, ``Criticality - Spent Fuel,''

specifies the minimum center-to-center distance, a Keff limit, and

the maximum enrichment for fuel in the spent fuel storage racks. These

requirements are also contained in STS Section 4.3.1.1, Items a, b, and

c. Section 5.6.2, ``Criticality - New Fuel,'' specifies the minimum

center-to-center distance, a keff limit, and a maximum enrichment

for fuel in the new fuel racks. These requirements are contained in STS

Section 4.3.1.2, Items a, b, c, and d. These sections meet Criteria 3,

e.g., geometries which, if altered, would have an immediate and

significant impact on safety. These requirements do not appear in other

sections of the TS. The STS language contains the same restrictions as

the Calvert Cliffs Design Features while introducing no new

requirements. BG&E proposes adopting the STS language and we will add

information on uncertainties of the referenced sections of the UFSAR.

Section 5.6.3 Drainage

This section in the Calvert Cliffs Design Features section and the

corresponding STS Section 4.3.2 does not meet any of the criteria. It

does not meet Criteria 3 in that it does not describe geometry or

materials of construction and because the requirements are contained in

another portion of the TS. BG&E proposes that this section be

eliminated from the Calvert Cliffs Design Features section and from the

STS.

Section 5.6.4 Capacity

Section 5.6.4, ``Capacity,'' states the maximum spent fuel storage

capacity. This meets Criteria 1 in that it limits the amount of special

nuclear material that may be stored on site. The Calvert Cliffs TS

language varies slightly from the STS language in that it makes clear

that the storage capacity limit applies to the combined storage pool

for Units 1 and 2. Therefore, we propose to retain the Calvert Cliffs

TS language.

Section 5.7 Component Cyclic or Transient Limits

This section is not included in the STS and does not meet any of

the criteria. We propose that it be eliminated from the Calvert Cliffs

Design Features section.

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. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed change to the Design Features section adopts

language from the Design Features section of the Standard Technical

Specifications (STS) for Combustion Engineering Plants (NUREG-1432,

September, 1992), based on the Commission's Final Policy Statement

on Technical Specifications Improvements for Nuclear Power Reactors

(July 16, 1993). Some requirements in the current Technical

Specifications have been eliminated or relocated to the UFSAR based

on the STS as evidence that the NRC no longer considers those

requirements to meet the criteria for Design Features in 10 CFR

50.36(c)(4). In some cases, we have proposed elimination of some

requirements in the STS and the Calvert Cliffs Technical

Specifications based on a determination that they have no legal or

regulatory basis.

We propose eliminating the ``Site'' sections, present in the

Calvert Cliffs and the STS Design Features sections, which contain

maps or descriptions of the site boundary and low population zone.

There are no legal or regulatory requirements for including this

information in the Technical Specifications and this information

will be relocated to the UFSAR. We propose adding a ``Site

Location'' section which contains a description of the Calvert

Cliffs location as required by the Atomic Energy Act of 1954 [,as

amended]. The ``Containment,'' ``Reactor Coolant System,''

``Meteorological Tower Location,'' and ``Component Cyclic or

Transients Limits'' sections do not meet the legal or regulatory

requirements for inclusion in the Design Features section, are not

included in the STS, and have been deleted. The information in these

sections is or will be contained in the UFSAR and will be controlled

under 10 CFR 50.59. The ``Control Element Assemblies'' and ``Fuel

Storage - Drainage'' sections are included in the STS and Calvert

Cliffs Technical Specifications but eliminated in this change as

these sections do not meet any of the legal or regulatory

requirements for the Design Features section. This information is

contained in the UFSAR and will be controlled under 10 CFR 50.59. We

propose adopting the STS wording for the ``Reactor Core'' and ``Fuel

Storage'' sections with no changes in the present limits or

controls. Some information in the Calvert Cliffs Technical

Specifications is not contained in the STS ``Reactor Core'' section.

This information is contained in the UFSAR and will be controlled

under 10 CFR 50.59.

All information eliminated from the Design Features section of

the Technical Specifications is or will be located in the UFSAR and

will be controlled under 10 CFR 50.59. The design and operation of

the plant have not changed. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The proposed change does not represent a change in the

configuration or operation of the plant. All information eliminated

from the Technical Specifications will continue to be controlled

under 10 CFR 50.59. All legal and regulatory requirements for the

Design Features section continue to be met. Therefore, the proposed

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

accident from any accident previously evaluated.

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

safety.

The proposed change does not represent a change in the

configuration or operation of the plant. All information eliminated

from the Technical Specifications will continue to be controlled

under 10 CFR 50.59. All legal and regulatory requirements for the

Design Features section continue to be met. Therefore, the proposed

change does not involve a significant reduction in a margin of

safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Robert A. Capra

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: December 10, 1993

Description of amendment request: The proposed amendment would

extend existing plant surveillance intervals to 24 months from 18

months. This is the third of three submittals and it changes specific

setpoints to accommodate a 24 month fuel cycle and provides a

justification for extending the surveillance interval for those

components and systems that are not related to instrument setpoint

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:

1. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

This submittal results in changes to various equipment

surveillance intervals and related instrument calibration

frequencies and setpoints.

The impact of lengthening the current 18 month interval to 24

months was evaluated and identified no significant system or

component degradation as a consequence of lengthening the interval

to 24 months; therefore, systems and components will continue to

perform their design function. In some cases, the 24 month interval

required setpoint changes to ensure instrument drift associated with

the extended interval would not result in exceeding an instrument's

acceptable setpoint tolerance. In other cases, justification of an

extended interval was not developed because the surveillance could

be performed on-line. In these cases, the ``once/cycle''

surveillance requirement is changed to the currently allowed 18

months.

The proposed changes were developed using the guidance provided

in Generic Letter 91-04 and Note 1 of Table 4.2.A through 4.2.G of

Pilgrim's technical specifications. The proposed changes do not

degrade the performance or increase the challenges to the associated

safety systems assumed to function in the accident analysis.

The impact of lengthening the current calibration/functional

test interval from 1 to 3 months for certain components was also

evaluated. The evaluation used the guidance in Generic Letter 91-04

and Note 1 of Table 4.2.A through 4.2.G of Pilgrim's technical

specifications. No significant system or component degradation was

identified as a consequence of lengthening the interval to 3 months.

The proposed changes do not affect the availability of equipment

or systems required to mitigate the consequences of an accident, and

do not affect the availability of redundant systems or equipment.

The plant will continue to operate within the limits specified in

the Core Operating Limits Report (COLR) and will continue to take

the same actions if setpoint limits are exceeded.

Therefore, both the proposed setpoint and non-setpoint changes

do not significantly increase the probability or consequences of an

accident previously evaluated.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously analyzed.

The proposed changes with one exception, do not add or remove

active components and, therefore, do not introduce failure

mechanisms of a different type than those previously evaluated. In

one case, the EDG breaker time delay relays will be replaced with

more accurate relays to ensure the specified time sequence for

starting and accepting the emergency load remains within

specification for the extended cycle. The replacement relays will be

similar in size, weight, voltage and temperature operating range as

those being replaced; therefore, the possibility of a new or

different kind of accident is not created. In addition, the

surveillance test requirements and the way surveillance tests are

performed will remain unchanged. Since the intended operation and

function of the analyzed systems do not change as a result of the

setpoint and non-setpoint analyses, no new initiators are introduced

capable of initiating an accident that would render these systems

unable to provide their required protection. Therefore, the proposed

changes do not create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in the

margin of safety.

Although the proposed Technical Specification changes will

result in an increase in the interval between surveillance tests,

the existing margins of safety are maintained through our proposed

setpoint revisions. The proposed setpoint changes either increase

the plant safety margin or maintain the existing margin and do not

significantly impact the availability, performance, or intended

function of the affected systems. In the case of non-setpoint

changes, evaluation of the affected systems indicates lengthening

the interval to 24 months does not have significant impact on

performance. Therefore, the assumptions in Pilgrim's accident

analyses are not impacted, and the proposed Technical Specification

changes do 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 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler

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, Illinois

Date of amendment request: March 26, 1993

Description of amendment request: The proposed amendment is a part

of Commonwealth Edison Company's (CECo's) Technical Specification

Upgrade Program (TSUP) to improve the quality of the current Technical

Specifications (TS) for Dresden and Quad Cities. The proposed amendment

would for both Dresden and Quad Cities, upgrade the requirements of

Section 3.9/4.9, ``Auxiliary Electrical Systems,'' to include operating

and shutdown Limiting Condition(s) for Operation (LCO) and Surveillance

Requirement(s) (SR) that are consistent with the Standard Technical

Specifications (STS) and later operating plant provisions. Within the

upgrade to Section 3.9/4.9, Emergency Diesel Generator (EDG)

reliability provisions are added to implement the recommendations of

Generic Letter(s) 84-15 and 91-09; Information Notice(s) 84-69 and 91-

62; and Regulatory Guide 1.9, draft Revision 3. Other Generic Letters

considered in the proposed TS include 83-26, 83-30, and 87-09.

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:

Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

In general, the proposed changes represent the conversion of

current requirement to a more generic format, or the addition of

requirements which are based on the current safety analyses.

Implementation of these changes will provide increased reliability

of equipment assumed to operate in the current safety analyses, or

provide continued assurance that specified parameters remain within

their acceptance limits, and as such, will not significantly

increase the probability or consequences of a previously evaluated

accident.

Some of the proposed changes represent minor curtailments of the

current requirements which are based on generic guidance or

previously approved provisions for other stations. These proposed

changes are consistent with the current safety analyses and have

been previously determined to represent sufficient requirements for

the assurance and reliability of equipment assumed to operate in the

safety analyses, or provide continued assurance that specified

parameters remain within their acceptance limits. As such, these

changes will not significantly increase the probability or

consequences of a previously evaluated accident.

A.C. Sources - Operating: The proposed modifications for Section

3/4.9.A, ``A.C. Sources - Operating'', administratively incorporate

the requirements of STS, where applicable to Dresden and Quad Cities

Stations. Most deviations from the STS requirements are based upon

generic guidance and other approved requirements at other sites.

Dresden and Quad Cities Station are retaining the current seven day

allowed outage time from their current specifications for loss of an

EDG. However, an additional verification of EDG operability has been

proposed for approximately midway through the seven day AOT. The

additional verification and additional details in the surveillances

will significantly improve the overall safety of both Dresden and

Quad Cities Station.

Both Dresden and Quad Cities Station's EDG history have shown

them to be very reliable. This can be demonstrated by the excellent

pass/fail rate observed during the monthly surveillance tests. As

stated previously, the addition of several new STS enhancements to

fuel storage and transfer requirements and other miscellaneous EDG

surveillances recommended by ASTM codes will improve EDG

reliability. Therefore, because the EDG's for Dresden and Quad

Cities Station do not act as accident initiators, the probability of

an accident previously evaluated for the sites is not increased by

the incorporation of the proposed requirements.

Other changes based upon STS guidance are more restrictive and

limit operation of the site with respect to all A.C. power sources.

A.C. power sources do not act as initiators of accidents. Therefore,

the probability of an accident previously evaluated for the sites is

also not increased by the incorporation of the STS requirements.

The consequences of any previously evaluated accidents are not

increased as more restrictions and limitations are added to the

current versions of both Dresden and Quad Cities specifications. The

retention of the seven day allowed outage times for the EDG's and

the offsite power sources does not increase the consequences of any

previously analyzed accident for both sites as these are the current

requirements. Therefore, the consequences of any previously

evaluated accident is not increased as a result of the proposed

changes.

A.C. Sources - Shutdown: The proposed modification for Section

3/4.9.B, ``A.C. Sources - Shutdown'', administratively incorporate

the requirements of STS where applicable to Dresden and Quad Cities

Station. The STS requirements add additional provisions not in the

current Technical Specifications for EDG fuel storage capability

that will reduce the consequences of a previously analyzed accident.

The probability of any previously evaluated accident is reduced

during shutdown by the additional STS restrictions proposed for fuel

handling type of activities. Therefore, the proposed changes do not

involve a significant increase in the probability or consequences of

any previously evaluated accident for Dresden and Quad Cities

Station.

D.C. Sources - Operating: The proposed modifications for Section

3/4.9.C, ``D.C. Sources - Operating'', incorporate the requirements

of STS where applicable for Dresden and Quad Cities Stations.

Dresden and Quad Cities are proposing to retain the current

provisions specified in the current version of Quad Cities Technical

Specifications that allow the 125 and 250 VDC systems to be out-of-

service for a period up to 72 hours. This 72-hour AOT may be

extended, when applied to the 125 VDC systems, for up to a maximum

period of 7 days with both units operating if the alternate 125 VDC

battery is operable. These changes introduce a difference when

compared to the STS requirements. However, the STS requirements as

applied to Dresden and Quad Cities battery systems would prove to be

overly burdensome requiring dual unit shutdowns to perform most

maintenance or testing activities. The additional batteries in the

design of the stations, with their surveillances, charger

requirements, and breaker verifications, compensate for any STS

deviations. The proposed requirements are comparable to the existing

requirements and AOTs for Quad Cities Station; therefore, the

consequences of any previously evaluated accident are not increased.

The proposed changes add additional surveillance requirements to the

D.C. systems at Dresden and Quad Cities to enhance their reliability

and operational readiness. This also ensures the consequences of any

previously evaluated accident are not increased. Because the D.C.

system is not assumed as an accident initiator, the probability of

any previously evaluated accident is not increased.

D.C. Sources - Shutdown: The proposed modifications for Section

3/4.9.D, ``D.C. Sources - Shutdown'', administratively incorporate

the requirements of STS where applicable to Dresden and Quad Cities

Station. The proposed changes add additional surveillance

requirements and more explicitly clarify the LCO's. The additional

provision for fuel handling type of activities reduces the

probability of previously evaluated accidents from occurring. The

additional surveillance activities also improve D.C. reliability and

thus, reduce the probability of D.C. system unavailability and

hence, reduce the consequences of previously evaluated accidents.

Because the D.C. system is not considered as an accident initiator,

the probability of any previously evaluated accident is not

increased.

Distribution - Operating: The proposed modifications for Section

3/4.9.E, ``Distribution - Operating'', administratively incorporate

the requirements of STS where applicable to Dresden and Quad Cities

Station. The proposed changes add additional surveillance

requirements and LCO's. The proposed requirements/actions for the

D.C. distribution system are retained to be consistent to the

proposed AOTs for the D.C. system. The STS requirements as applied

to the Dresden/Quad Cities D.C. distribution system would prove to

be overly burdensome, requiring plant shutdowns to perform

maintenance or testing activities. The additional distribution

system surveillance and LCO's compensate for the STS deviations. The

proposed requirements are comparable to the existing requirements

and AOTs for Quad Cities Station; therefore, the consequences of any

previously evaluated accident are not increased. The additional

surveillances and STS-type requirements ensures the reliability and

operational readiness of the Distribution System and ensures the

consequences of any previously evaluated accident are not increased.

Because the Distribution System is not assumed as an accident

initiator, the probability of any previously evaluated accident is

not increased.

Distribution - Shutdown: The proposed modification for Section

3/4.9.F, ``Distribution - Shutdown'', administratively incorporate

the requirements of STS where applicable for Dresden and Quad Cities

Station. The proposed changes add additional requirements that

ensure the consequences and the probability of any previously

evaluated accident are not increased.

RPS Power Monitoring: The proposed modifications for Section 3/

4.9.G, ``RPS Power Monitoring''. administratively incorporate the

requirements of STS for Dresden and Quad Cities Station. The

proposed changes add additional requirements for Dresden and clarify

the existing requirements at Quad Cities. Therefore, the

consequences and the probability of any previously evaluated

accident are not increased.

Create the possibility of a new or different kind of accident

from any previously evaluated because:

In general, the proposed changes represent the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analyses. Others

represent minor curtailments of the current requirements which are

based on generic guidance or previously approved provisions for

other stations. These changes do not involve revisions to the design

of the station. Some of the changes may involve revision in the

operation of the station; however, these provide additional

restrictions which are in accordance with the current safety

analyses, or are to provide for additional testing or surveillances

which will not introduce new failure mechanisms beyond those already

considered in the current safety analyses. The retention of the

current AOTs for EDGs, offsite power sources, and DC systems

maintain the existing assumptions from the current accident

analyses; therefore, these changes will not create the possibility

of a new or different kind of accident from any previously

evaluated.

The proposed changes for Dresden and Quad Cities Station's

Technical Specification Section 3/4.9 are based on STS guidelines or

later operating BWR plants' NRC accepted changes. These proposed

changes have been reviewed for acceptability at the Dresden and Quad

Cities Nuclear Power Stations considering similarity of system or

component design versus the STS of later operating BWRs. No new

modes of operation are introduced by the proposed changes,

considering the acceptable operational modes in present

specifications, the STS, or later operating BWRs. Surveillance

requirements are changed to reflect improvements in technique,

frequency of performance or operating experience at later plants.

Proposed changes to action statements in many places add

requirements that are not in the present technical specifications or

adopt requirements that have been used successfully at other

operating BWRs with designs similar to Dresden and Quad Cities. The

proposed changes maintain at least the present level of operability.

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

new or different kind of accident from any previously evaluated.

Involve a significant reduction in the margin of safety because:

In general, the proposed changes represent the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analyses. Others

represent minor curtailments of the current requirements which are

based on generic guidance or previously approved provisions for

other stations. Some of the later individual items may introduce

minor reductions in the margin of safety when compared to the

current requirements. However, other individual changes are the

adoption of new requirements which will provide significant

enhancement of the reliability of the equipment assumed to operate

in the safety analyses, or provide enhanced assurance that specified

parameters remain with their acceptance limits. These enhancements

compensate for the individual minor reductions, such that taken

together, the proposed changes will not significantly reduce the

margin of safety.

The proposed changes to the Technical Specification Section 3/

4.9 implement present requirements, or the intent of present

requirements in accordance with the guidelines set forth in the STS.

The proposed changes are intended to improve readability, usability,

and the understanding of technical specification requirements while

maintaining acceptable levels of safe operation. The proposed

changes have been evaluated and found to be acceptable for use at

Dresden and Quad Cities based on system design, safety analyses

requirements and operational performance. The retention of the

current AOTs for EDGs, offsite power sources, and DC systems

maintain the existing assumptions from the current accident

analyses. Since the proposed changes are based on NRC accepted

provisions at other operating plants that are applicable at Dresden

and Quad Cities and maintain necessary levels of system, component

or parameter readability, the proposed changes do not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: for Dresden, the Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450, and for Quad

Cities, the Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois

61021

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

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

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

Date of amendment request: November 11, 1993

Description of amendment request: The proposed amendments would

consolidate the Quality Verification Department with the Nuclear

Generation Department and realign the Nuclear Safety Review Board such

that it reports to the Senior Vice-President of the Nuclear Generation

Department.

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 amendments do not involve a significant increase in the

probability or consequences of an accident previously evaluated.]

The proposed revisions to consolidate the Quality Verification

Department with the Nuclear Generation Department and realign the

NSRB [Nuclear Safety Review Board] such that it reports to the

Senior Nuclear Officer, change the reference from Semiannual to

Annual, change the reference from group to division, delete titles

of persons designated to approve modifications, clarify the

responsibilities of the Safety Assurance Manager, and delete the

requirement to perform an annual independent Fire Protection Audit

will not involve a significant increase in the probability or

consequences of an accident previously evaluated because the changes

do not have any impact upon the design or operation of any plant

systems or components.

[2. The amendments do not create the possibility of a new or

different kind of accident from any accident previously evaluated.]

The proposed revisions will not create the possibility of a new

or different kind of accident from any previously evaluated because

the changes are administrative in nature and operation of Catawba,

McGuire, and Oconee Nuclear Stations in accordance with these TS

[technical specifications] will not create any failure modes not

bounded by previously evaluated accidents.

[3. The amendments do not involve a significant reduction in a

margin of safety.]

The proposed revisions will not involve a reduction in a margin

of safety because they are administrative in nature.

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: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Loren R. Plisco, Acting

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 6, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to add provisions to allow

repair of steam generator tubes by the sleeving process.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change will allow the use of specific steam

generator tubing sleeves to repair Waterford 3 steam generator tubes

which exhibit degradation and can be sleeved (in the tube sheet

crevice area and egg crate supports). The technical specification

change is proposed to reference the following reports [following NRC

approval], which qualify the use of steam generator tube sleeves as

an alternative to tube plugging,

Combustion Engineering Report CEN-605-P, ``Waterford 3 Steam

Generator Tube Repair Using Leak Tight Sleeves'', Revision 00-P,

dated December, 1992.

Westinghouse Report WCAP-13698, ``Laser Welded Sleeves For 3/4

Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam

Generators'', Revision 1, dated May, 1993.

Babcock & Wilcox Report 51-1223750-00, ``BWNS Kinetic Sleeve

Design For CE SGs with 0.048'' Wall Tubes'', Revision 00, dated June

29, 1993.

These reports demonstrate that the repair of degraded steam

generator tubes using tube sleeves will result in tube bundle

integrity consistent with the original design basis.

Sleeve design, materials, and joints were designed to the

applicable ASME Boiler and Pressure Vessel Codes. Extensive analyses

and testing were performed to demonstrate the adequacy of the tube

sleeves. The analyses were performed using design and operating

transient parameters which enveloped loads imposed during normal

operating, upset and accident conditions. Mechanical testing was

performed to demonstrate leak resistance and joint strength,

including fatigue resistance.

Corrosion testing was also performed to assess the corrosion

resistance of the sleeve and joint. Based upon the results of the

analytical and test programs described in detail in the above

mentioned reports, these tube sleeves meet or exceed all the

established design and operating criteria.

Utilization of tube sleeves not only reduces the risk of primary

to secondary leakage in the steam generator, but can also provide

for more margin in the safety analysis. Sleeving a tube results in a

primary flow reduction which has no significant effect on the steam

generator performance with respect to heat transfer or system flow

resistance and pressure drop. The cumulative impact of multiple

sleeved tubes has been evaluated to ensure the effects remain within

the design bases. The installation of tube sleeves can be

accomplished within the tube plugging analysis.

Based on the extensive analysis and test program performed and

the ability to monitor and remove degraded sleeves from service,

this change does not significantly increase the probability or

consequences of an accident previously evaluated.

A sleeved tube performs the same function in the same passive

manner as the unsleeved tube. Tube sleeves are designed, qualified,

and maintained under the stress and pressure limits of ASME Section

III and Regulatory Guide 1.121. Eddy current testing is performed

following installation of each sleeve in order to verify the proper

installation of the tube sleeve and to obtain baseline eddy current

data. This baseline data is used to monitor any subsequent

degradation.

Therefore, the use of tube sleeves does not create the

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

previously evaluated.

Steam generator tube integrity is maintained under the same

limits for sleeved tubes as for unsleeved tubes in accordance with

ASME Section III and Regulatory Guide 1.1.21. The degradation limit

at which a tube is considered inoperable remains unchanged and is

detectable for sleeves as well as tubes. The technical

specifications continue to require monitoring and restriction of

primary to secondary system leakage through the steam generators,

such that there remains reasonable assurance that a significant

increase in leakage, due to failure of a sleeved (or unsleeved)

tube, will be detected. The slight reduction in reactor coolant

system flow, due to sleeving, is considered to have an insignificant

impact on steam generator operation during normal operation and

accident conditions and is clearly bounded by tube plugging

evaluations. The technical specifications will continue to contain

reporting requirements for tubes which have had their degradation

spanned regardless of whether the tube is plugged or sleeved.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 14, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to revise the Azimuthal Power

Tilt limit from less than or equal to 0.10 (10%) to less than or equal

to 0.03 (3%) and to revise the action statement for control element

assembly (CEA) misalignment to allow 24 hours to restore the tilt to

less than 3%.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The additional time for recovery from a CEA misalignment is

acceptable for the following reasons:

Consistent with the safety analyses, this TS, places a limit on

tilt for steady state operation as an initial condition for the

safety analyses. It is not a limit to be applied during transients.

This is because accident analyses are initiated from steady state

conditions and are not required to assume a core power distribution

transient simultaneous with or immediately prior to the accident.

This would in effect be two accidents occurring simultaneously.

The probability of having an accident immediately after a CEA

drop during the 24 hour period allowed for tilt to be restored to

less than 3% is very low.

Technical Specification 3/4.1.3 addresses the CEA misalignments

and requires a 30% power reduction within one hour if the CEA cannot

be restored to its proper position.

Reducing power as required in TS 3.2.3 action b.2 will tend to

increase the azimuthal tilt, making the transient worse. A lower

power reduces the rate at which xenon near the dropped CEA can be

burned out. Maintaining power will quicken the process and keep the

tilt as low as possible.

The additional time is only allowed for a confirmed CEA

misalignment which operators have procedures to respond to. The

change in tilt is expected and is not indicative of anomalous core

power distribution behavior that might require more immediate

action.

This change conservatively reduces the Azimuthal Power Tilt

technical specification limit to agree with the assumptions used in

the safety analysis. The lower tilt represents a more even power

distribution in the core. A CEA drop event may temporarily cause the

azimuthal power tilt to exceed the 3% limit. However, for the

reasons identified above and since the probability of having another

event within the 24 hours allowed for recovery after the CEA drop is

extremely low, this change does not involve a significant increase

in the probability or consequences of any accident.

The change in technical specification limit on tilt does not

involve any change to any equipment or the manner in which the plant

will be operated. This change will further restrict unevenness in

the core power distribution. Therefore, this change does not create

the possibility of a new or different kind of accident previously

evaluated.

The proposed change incorporates an Azimuthal Power Tilt

technical specification limit to agree with the assumptions used in

the safety analysis. Implementation of this change will preserve the

margin of safety and be consistent with the safety analyses.

Therefore, this change does not involve a significant reduction in

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 14, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) by removing the reactor

vessel material specimen withdrawal schedule and by updating the

reactor coolant system pressure-temperature (P-T) curves. The specimen

withdrawal schedule will be relocated to the Updated Final Safety

Analysis Report (UFSAR).

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:

Although the Reactor Vessel material specimens withdrawal

schedule will be removed from the Technical Specifications, the

Technical Specifications bases will continue to provide background

information on the use of the data obtained from material specimens.

Also, updates to the schedule will continue to be submitted to the

NRC for approval prior to implementation.

Operating the plant in accordance with the new, updated P-T

Curves will assure preserving the structural integrity of the

reactor vessel over the life of the plant. The pressure and

temperature limits were developed in accordance with 10 CFR [Part]

50 Appendix G requirements.

Removing the requirements associated with the previous exemption

to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative

change.

Therefore, the proposed changes will not significantly increase

the probability or consequences of any accident previously

evaluated.

Removal of the Reactor Vessel material specimen schedule from

the Technical Specifications has no impact on accidents at the

plant. Updates to the schedule will still be required to be

submitted to the NRC prior to implementation per Section II.B.3 of

Appendix H to 10 CFR Part 50.

Also, updates to the P-T Curves will not create a new or

different type [of] accident. The reactor vessel beltline P-T limits

were revised applying the general guidance of the ASME Code,

Appendix G procedures with the necessary margins of safety for

heatup, cooldown and inservice hydro test conditions.

The change to TS 4.4.8.1.2 items a & b is a purely

administrative.

Therefore, the proposed changes will not create the possibility

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

evaluated.

Removal of the schedule for Reactor Vessel material specimen

withdrawal from the Technical Specifications does not impact the

margin of safety. The schedule will continue to receive NRC review

and approval prior to implementation of updates to the schedule.

Updates to the P-T Curves are provided to preserve the margin to

[sic] safety to assure that when stressed under operating,

maintenance and testing the boundary behaves in a non-brittle manner

and the probability of rapidly propagating fracture is minimized.

The change to TS 4.4.8.1.2 items a & b is a purely

administrative.

Therefore, the proposed changes 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 New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

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

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: November 30, 1993

Description of amendment request: The purpose of the request is to

change the plant Technical Specifications (TS) by removing the

protective and maximum allowable setpoint limits for axial power

imbalance and the trip setpoint for nuclear overpower based on reactor

coolant system (RCS) flow (flux-to-flow) from the TS and relocating

them to the existing TMI-1 Core Operating Limits Report (COLR). The

proposed change is in accordance with Generic Letter 88-16 guidance

with regard to placing cycle-dependent parameters into the COLR and the

NRC-approved Babcock and Wilcox Fuel Company (BWFC) Topical Report BAW-

10179P-A, ``Safety Criteria and Methodology for Acceptable Cycle Reload

Analyses.'' The TMI-1 Cycle 10 COLR, submitted to the NRC on November

7, 1993, includes these protective and maximum allowable setpoint

limits and nuclear overpower trip setpoints to support this Technical

Specifications change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated. The proposed amendment relocates protective and maximum

allowable setpoint limits from Technical Specifications, and design

nuclear power peaking factors and the maximum allowable local linear

heat rate limit from Technical Specification Bases, to the TMI-1

Core Operating Limits Report in accordance with NRC-approved Topical

Report BAW-10179P-A. The proposed amendment provides continued

control of the values of these limits and assures these values

remain consistent with all applicable limits of the safety analyses

addressed in the TMI-1 FSAR [Final Safety Analysis Report]. The

Technical Specifications retain the requirement to maintain the

plant within the appropriate bounds of these limits. Therefore, the

proposed amendment has no effect on the probability of occurrence or

consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated. The

proposed amendment relocates protective and maximum allowable

setpoint limits, design nuclear power peaking factors and maximum

allowable local linear heat rate limit to the TMI-1 Core Operating

Limits Report. The Technical Specifications retain the requirement

to maintain the plant within the appropriate bounds of these limits.

Therefore, the proposed amendment has no effect on the possibility

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

previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment provides continued control of the

values of these limits and assures these values remain consistent

with all applicable limits of the safety analyses addressed in the

TMI-1 FSAR. Therefore, it is concluded that operation of the

facility in accordance with the proposed amendment does not involve

a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz

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

Point Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: December 14, 1993

Description of amendment request: The amendment would revise

Technical Specification (TS) 3/4.8.2, ``DC Sources,'' to delete two

notes that indicate that two 125-volt full capacity battery chargers

are required when the Uninterruptible Power Supply is powered by its

backup DC power supply. These notes apply to the Divisions I and II DC

sources during operating and shutdown conditions. The licensee has

determined that only one battery charger is required to meet current

design requirements and the criteria delineated in Regulatory Guide

1.32, ``Criteria for Safety-Related Electric Power Systems for Nuclear

Power Plants.'' The amendment would also revise TS 3/4.8.2 to increase

the minimum allowable electrolyte temperature for the 125-volt

batteries from 60 deg.F to 65 deg.F. This proposed change would

establish consistency between the TSs, the Updated Safety Analysis

Report, and applicable battery capacity calculations. The amendment

would also make administrative changes to TS 3/4.8.4, ``Electrical

Equipment Protective Devices,'' and the TS Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The existing ``notes'' at the bottom of pages 3/4 8-14 and 3/4

8-19 indicate that two (2) 125-volt full capacity chargers are

required when the Uninterruptible Power Supply is powered by its

backup DC power supply. These ``notes'' were based on an overly

conservative calculation which determined that both chargers were

required to supply adequate power to connected loads including an

Uninterruptible Power Supply. More recent calculations indicate that

one (1) charger is adequate to supply power to connected loads and

an Uninterruptible Power Supply. Based on these more recent

calculations, Niagara Mohawk proposes to delete these ``notes.''

Because it has been determined adequate power will be provided to

connected loads with one (1) charger, deletion of these ``notes''

will not affect the reliability of connected loads nor their ability

to perform their intended function. Therefore, this change will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Existing Technical Specification 4.8.2.1 requires that the 125-

volt batteries be demonstrated to be operable by verifying that the

average electrolyte temperature of one out of five connected cells

is above 60 deg.F. The Nine Mile Point Unit 2 Updated Safety

Analysis Report and current battery capacity calculations assume a

battery electrolyte temperature of at least 65 deg.F. The change

from 60 deg.F to 65 deg.F is conservative in that battery capacity

is increased. Increasing the capacity of the batteries will not

adversely affect the reliability of connected loads nor their

ability to perform their intended function. Therefore, this change

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The changes made to Technical Specification 3/4.8.4 and to the

Bases of Specifications 3/4.6.3 and 3/4.8.4 are administrative

changes and do not affect plant systems or operation. Accordingly

these changes will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The function of the battery chargers is to provide adequate

power to connected loads. Since it has been determined that one (1)

battery charger is sufficient to provide adequate power, deletion of

these ``notes'' which require that two (2) chargers be available

does not affect the capability of the chargers to perform their

function. The proposal to change the required electrolyte

temperature from 60 deg.F to 65 deg.F increases the capacity of the

batteries and therefore improves the capability of the batteries to

perform their intended function. The remaining changes are

administrative changes and do not affect plant systems or operation.

The proposed changes do not introduce any new accident

precursors and do not involve any physical alterations to plant

configurations which could initiate a new or different kind of

accident. The changes do not adversely affect the design or

performance characteristics of the batteries, battery chargers or

connected loads. The proposed change to increase the required

electrolyte temperature will increase battery capacity. Therefore,

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

different kind of accident from any previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

Niagara Mohawk proposes to delete the existing ``notes'' that

indicate that two (2) 125-volt full capacity chargers are required

when the Uninterruptible Power Supply is powered by its backup DC

power supply. Niagara Mohawk engineering has determined that one (1)

battery charger is adequate to meet the maximum DC load demands

including the Uninterruptible Power Supply. The proposed change to

increase the required electrolyte temperature will increase battery

capacity. The remaining changes are administrative.

These changes will not adversely affect the design or

performance characteristics of the batteries, battery chargers, or

connected loads nor will they affect the capability of the

batteries, battery chargers, or connected loads to perform their

intended function. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

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

NRC Project Director: Robert A. Capra

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

Power Plant, Unit 3, Humboldt County, California

Date of application for amendment: July 7, 1993 (Reference LAR 93-

01)

Brief description of amendment: The proposed amendment would revise

the Technical Specifications (TS) for the Humboldt Bay Power Plant Unit

No. 3. This proposed revision would change TS VII.H.3, ``Semiannual

Radioactive Effluent Release Report,'' to extend the reporting period

from semiannually to annually and to change the report submission date

from 60 days after January 1 and July 1 of each year to before April 1

of each year.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed TS changes are administrative in nature. The

proposed changes to TS VII.H.3 are consistent with 10 CFR 50.36a

report requirements. The proposed changes do not affect accident

evaluations. The proposed changes are administrative in nature,

should result in improved administrative practices, and do not

affect plant operations.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature, do not result

in physical alterations or changes to the operation of the plant,

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

performs its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

These administrative changes do not alter the basic regulatory

requirements and do not affect any safety analyses.

The proposed change to TS VII.H.3. does not alter any

administrative controls over radioactive effluent, nor does the

proposed change involve any physical alterations to the plant with

respect to radioactive effluents. Therefore, the proposed change

would not affect the meaning, application, and function of the TS

requirements.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Humboldt County Library, 636 F

Street, Eureka, California 95501

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

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

NRC Branch Chief: John H. Austin

Portland General Electric Company, et al., Docket No. 50-344,

Trojan Nuclear Plant, Columbia County, Oregon

Date of amendment request: January 27, 1993

Description of amendment request: The proposed amendment, by

Portland General Electric Company, PGE or the licensee, would change

the Trojan Nuclear Plant (Trojan) Appendix A Technical Specifications

to reflect the permanently defueled status of the facility. The

permanent cessation of power generation at Trojan and the May 5, 1993

amendment to the license which granted the licensee a Possession Only

License for the facility has rendered many of the existing provisions

of the current Appendix A Technical Specifications inappropriate. PGE

has developed Permanently Defueled Technical Specifications (PDTS) for

Trojan using NUREG-1431, ``Standard Technical Specifications,

Westinghouse Plants,'' as a basis for the PDTS scope and format.

Basis for proposed no significant hazards consideration

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

provided an analysis of the issue of no significant hazards

consideration based upon the following:

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 shows that the worst case design basis

accident for this plant, in its permanently shutdown defueled state,

is a postulated spent fuel handling accident at the Trojan facility.

The licensee has also identified a second postulated design basis

accident scenario, a loss of forced cooling to the spent fuel pool.

Other Trojan Final Safety Analysis Report (FSAR) accident scenarios

addressed in Chapter 15 are no longer applicable to Trojan in the

permanently defueled mode. The proposed amendment does not lessen

any of the requirements associated with handling spent fuel and

therefore, the probability of a fuel handling accident occurring is

unchanged. The licensee has analyzed the a loss of forced spent fuel

pool cooling accident and has shown that this scenario would not

result in a radiological release. The proposed amendment does not

change the consequences of the accident since it does not affect the

magnitude, detection, or mitigation of either accident scenario.

Additionally, the ability of the spent fuel pool to withstand other

applicable FSAR events, natural phenomena, and fires is either

unchanged from the existing licensing basis or is improved during

the permanently defueled condition.

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.

Maintaining the permanently defueled facility in accordance with

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

of accident from any previously considered. Most of the existing

plant systems and functions will not be operational in the

permanently defueled condition since power operations are prohibited

and all of the fuel at Trojan is stored in the spent fuel pool.

However, all structures, systems and components that are necessary

for safe fuel handling and storage activities will be maintained

operable during the permanently defueled condition. The proposed

PDTS provide operation and surveillance requirements and

administrative controls which are sufficient to ensure that the

required structures, systems and components will be maintained

operable in the permanently defueled condition.

3. Operation of the facility in accordance with the proposed

amendment does not involve a significant reduction in a margin of

safety.

The proposed PDTS are sufficient to ensure no reduction in a

margin of safety, in part, because of the reduced range of design

basis accidents against which the facility must be protected now

that the facility is prohibited from power operations and is

permanently defueled. Only a fuel handling accident or a loss of

forced cooling to the spent fuel pool are relevant during the

permanently defueled condition. The margins of safety for both of

these accidents will remain the same or improve by maintaining the

facility in accordance with the proposed PDTS. None of the other

Chapter 15 FSAR accidents are applicable since power operations are

prohibited and the facility is permanently defueled. Additionally,

the margins of safety for other applicable FSAR events, natural

phenomena, and fires are either unchanged from the existing

licensing basis or are improved during the permanently defueled

condition.

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: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensees: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.

NRR Project Director: Seymour H. Weiss

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

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

Date of amendment request: December 20, 1993

Description of amendment request: The licensee has requested an

amendment to the Technical Specifications (TS) to revise Section 3.3.D

(Weld Channel and Penetration Pressurization System) to allow certain

portions of the Weld Channel Pressurization System (WCPS) to be

disconnected if they are determined to be inoperable and not

practicably accessible for repair. The WCPS continuously pressurizes

channels over welds in the steel liner of the containment building. To

be disconnected, an inoperable portion of the WCPS must be covered by

concrete such that repairs would involve removal of part of the

containment structure or the inoperable portion of the system is

located behind plant equipment inside the containment building such

that repairs would involve relocation of the equipment. The licensee

has requested this TS amendment since one portion of the WCPS has

become inoperable and, since it is buried in concrete below the

containment floor, cannot be practically repaired. In addition,

administrative changes would be made to TS Section 3.3.B (Containment

Cooling and Heat Removal) and TS Section 3.3.E (Component Cooling

System) to correct typographical errors. Specifically, TS Section

3.3.B.3.b references 3.3.A.3 when is should reference 3.3.B.1 and TS

Section 3.3.E.3.b references 3.3.A.3 when it should reference 3.3.E.1.

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:

Consistent with the criteria of 10 CFR 50.92, the enclosed

application is judged to involve no significant hazards based on the

following information:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The consequences of an accident previously

evaluated would not be affected by the disconnection of portions of

the WCPS [Weld Channel Pressurization System] because the accident

analyses do not assume the operation of any portion of the WC & PPS

[Weld Channel & Penetration Pressurization System]. Additionally,

operation of the WC & PPS is not taken credit for in any offsite

accident dose calculations. The probability of an accident

previously evaluated would not be increased because the

disconnection of any portion of the WCPS could not initiate an

accident. The administrative changes correct errors in the technical

specifications and technical specification bases. These

administrative changes have no affect on the probability or

consequences of an accident previously evaluated.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed license amendment does not create the possibility

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

evaluated. The allowance for the disconnection of sections of the

WCPS will allow the Authority [Power Authority of the State of New

York] to avoid repairs that can potentially degrade containment

integrity or the condition of vital equipment. The proposed license

amendment does not create the possibility of a new accident because

the disconnection of any portion of the WCPS could not initiate an

accident. The administrative changes correct errors in the technical

specifications and technical specification bases. These

administrative changes can not [cannot] create the possibility of a

new or different kind of accident from any accident previously

evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed amendment would not involve a significant reduction

in a margin of safety. The allowance for the disconnection of

sections of the WCPS will allow the Authority to avoid repairs that

can potentially degrade containment integrity or the condition of

vital equipment. The WC & PPS will still provide continuous

pressurization and monitoring of leak-tightness for the zones

incorporated into the containment penetrations and for at lease 80%

of the channels over the welds in the steel inner of the containment

building. The WC & PPS will continue to provide assurance that the

containment leak rate in the event of an accident is lower than that

assumed in the accident analyses because the accident analyses do

not assume that any section of the WC & PPS is operating. The

administrative changes correct errors in the technical

specifications and technical specification bases. These

administrative changes have no affect on 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 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 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Robert A. Capra

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: December 8, 1993. The December 8, 1993

request supersedes an earlier request dated November 17, 1992, which

was previously noticed (58 FR 52994). This notice supersedes the

previous notice.

Description of amendment request: The proposed amendments would

revise the Technical Specifications for Salem Units 1 and 2 to

incorporate the guidance provided by the staff in Generic Letter 90-06

(GL 90-06) as follows:

1. Specification 3/4.4.3 and 3/4.4.5 for Salem 1 and Salem 2,

respectively, ``RELIEF VALVES'', will incorporate the guidance of GL

90-06 with the following exceptions:

a. The surveillance requirement to test the emergency power supply

for the power operated relief valves (PORVs) and block valves has not

been incorporated. The PORVs and block valves are powered from the

emergency busses.

b. The entry conditions for one or both PORVs inoperable will not

be based on excessive seat leakage alone. Entry conditions will be

based on the capability of the PORV to be manually cycled consistent

with the Action Statements contained in NUREG-1431, Standard Technical

Specifications for Westinghouse Plants.

c. With both PORVs inoperable in Modes 1, 2, or 3 and not capable

of being manually cycled, or both block valves are inoperable, an

allowed outage time of 6 hours to restore on block valve or PORV to

operable status has been requested.

2. Specification 3.4.9.3 and 3.4.10.3 for Salem 1 and Salem 2,

respectively, ``OVERPRESSURE PROTECTION SYSTEMS'' will incorporate the

guidance of GL 90-06.

3. In addition to the guidance provided by GL 90-06, the following

changes have also been proposed.

a. For Salem 1, the reference to the specific American Society of

Mechanical Engineers (ASME) valve category would be deleted from

Specification 3.4.9.3.1.

b. Specification 3.5.3. ``ECCS SUBSYSTEMS - TAVE 131 and its transient

dose equivalent I131 reactor coolant specific activity will be

reduced by a factor of 4 in order to increase the allowable leakage in

the event of a steam line break.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Operation of Farley Unit 1 in accordance with the proposed

license amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Testing of model boiler specimens for free standing tubes at

room temperature conditions show burst pressures as high as

approximately 5000 psi [per square inch] for indications of outer

diameter stress corrosion cracking with voltage measurements as high

as 26.5 volts. Burst testing performed on pulled tubes with up to

7.5 volt indications show burst pressures in excess of 5900 psi at

room temperature. As stated earlier, tube burst criteria are

inherently satisfied during normal operating conditions by the

presence of the tube support plate. Furthermore, correcting for the

effects of temperature on material properties and minimum strength

levels (as the burst testing was done at room temperature), tube

burst capability significantly exceeds the R.G. [Regulatory Guide]

1.121 criterion requiring the maintenance of a margin of 1.43 times

the steam line break pressure differential on tube burst if through-

wall cracks are present without regard to the presence of the tube

support plate. Based on the existing data base this criterion is

satisfied with bobbin coil indications with signal amplitudes over

twice the 2.0 volt interim repair criteria, regardless of the

indicated depth measurement. This structural limit is based on a

lower 95 [percent] confidence level limit of the data. The 2.0 volt

criteria provides an extremely conservative margin of safety to the

structural limit considering expected growth rates of outside

diameter stress corrosion cracking at Farley. Alternate crack

morphologies can correspond to a voltage so that a unique crack

length is not defined by a burst pressure to voltage correlation.

However, relative to expected leakage during normal operating

conditions, no field leakage has been reported from tubes with

indications with a voltage level of under 7.7 volts for a 3/4 inch

tube with a 10 volt correlation to 7/8 inch tubing (as compared to

the 2.0 volt proposed interim tube repair limit). Thus, the proposed

amendment does not involve a significant increase in the probability

or consequences of an accident.

Relative to the expected leakage during accident condition

loadings, the accidents that are affected by primary-to-secondary

leakage and steam release to the environment are Loss of External

Electrical Load and/or Turbine Trip, Loss of All AC Power to Station

Auxiliaries, Major Secondary System Pipe Failure, Steam Generator

Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a

Control Rod Drive Mechanism Housing. Of these, the Major Secondary

System Pipe Failure is the most limiting for Farley in considering

the potential for off-site doses. The offsite dose analyses for the

other events which model primary-to secondary leakage and steam

release from the secondary side to the environment assume that the

secondary side remains intact. The steam generator tubes are not

subjected to a sustained increase in differential pressure, as is

the case following a steam line break event. This increase in

differential pressure is responsible for the postulated increase in

leakage and associated offsite doses following a steam line break

event. In addition, the steam line break event results in a bypass

of containment for steam generator leakage. Upon implementation of

the interim repair criteria, it must be verified that the expected

distribution of cracking indications at the tube support plate

intersections are such that primary-to-secondary leakage would

result in site boundary dose within the current licensing basis.

Data indicate that a threshold voltage of 2.8 volts would result in

through-wall cracks long enough to leak at steam line break

conditions. Application of the proposed repair criteria requires

that the current distribution of a number of indications versus

voltage be obtained during the refueling outages. The current

voltage is then combined with the rate of change in voltage

measurement and a voltage measurement uncertainty to establish an

end of cycle voltage distribution and, thus, leak rate during steam

line break pressure differential. The leak rate during a steam line

break is further increased by a factor related to the probability of

detection of the flaws. If it is found that the potential steam line

break leakage for degraded intersections planned to be left in

service coupled with the reduced specific activity levels allowed

result in radiological consequences outside the current licensing

basis, then additional tubes will be plugged or repaired to reduce

steam line break leakage potential to within the acceptance limit.

Thus, the consequences of the most limiting design basis accident

are constrained to present licensing basis limits.

2) The proposed license amendment does not create the

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

previously evaluated.

Implementation of the proposed interim tube support plate

elevation steam generator tube repair criteria does not introduce

any significant changes to the plant design basis. Use of the

criteria does not provide a mechanism which could result in an

accident outside of the region of the tube support plate elevations.

Neither a single or multiple tube rupture event would be expected in

a steam generator in which the repair criteria has been applied

(during all plant conditions). The bobbin probe signal amplitude

repair criteria is established such that operational leakage of

excessive leakage during a postulated steam line break condition is

not anticipated. Southern Nuclear has previously implemented a

maximum leakage rate limit of 140 gpd [gallons per day] per steam

generator on Unit 1. The R.G. 1.121 criterion for establishing

operational leakage rate limits that require plant shutdown are

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

crack before potential tube rupture. The 140 gpd limit provides for

leakage detection and plant shutdown in the event of the occurrence

of an unexpected single crack resulting in leakage that is

associated with the longest permissible crack length. R.G. 1.121

acceptance criteria for establishing operating leakage limits are

based on leak-before-break considerations such that plant shutdown

is initiated if the leakage associated with the longest permissible

crack is exceeded. The longest permissible crack is the length that

provides a factor of safety of 1.43 against bursting at steam line

break pressure differential. A voltage amplitude approximately 9

volts for typical outside diameter stress corrosion cracking

corresponds to meeting this tube burst requirement at the 95

[percent] prediction interval on the burst correlation. Alternate

crack morphologies can correspond to a voltage so that a unique

crack length is not defined by the burst pressure versus voltage

correlation. Consequently, typical burst pressure versus through-

wall crack length correlations are used below to define the

``longest permissible crack'' for evaluating operating leakage

limits.

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

at 1.43 times steam line break pressure differential and steam line

break conditions are about 0.53 inch and 0.84 inch, respectively.

Normal leakage for these crack lengths would range from about 0.4

gallons per minute to 4.5 gallons per minute respectively while

lower 95 [percent] confidence level leak rates would range from

about 0.06 gallons per minute to 0.6 gallons per minute,

respectively.

An operating leak rate of 140 gpd per steam generator has been

implemented on Unit 1. This leakage limit provides for detection of

0.4 inch long cracks at nominal leak rates and 0.6 inch long cracks

at the lower 95 [percent] confidence level and for three times

normal operating pressure differential at less than nominal leak

rates.

Based on the above, the implementation of interim plugging

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

of accident from any previously evaluated.

3) The proposed license amendment does not involve a significant

reduction in margin of safety.

The use of the interim tube support plate elevation repair

criteria is demonstrated to maintain steam generator tube integrity

commensurate with the requirements of R.G. 1.121. R.G. 1.121

describes a method acceptable to the NRC staff for meeting [General

Design Criteria] 2, 14, 15, 31, and 32 by reducing the probability

of the consequences of steam generator tube rupture. This is

accomplished by determining the limiting conditions of degradation

of steam generator tubing, as established by inservice inspection,

for which tubes with unacceptable cracking should be removed from

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

case conditions, the occurrence of outside diameter stress corrosion

cracking at the tube support plate elevations is not expected to

lead to a steam generator tube rupture event during normal or

faulted plant conditions. The most limiting effect would be a

possible increase in leakage during a steam line break event.

Excessive leakage during a steam line break event, however, is

precluded by verifying that, once the criteria are applied, the

expected end of cycle distribution of crack indications at the tube

support plate elevations would result in minimal, and acceptable

primary to secondary leakage during the event and hence help to

demonstrate radiological conditions are less than an appropriate

fraction of the 10 CFR [Part] 100 guideline.

The margin to burst for the tubes using the interim repair

criteria is comparable to that currently provided by existing

technical specifications.

In addressing the combined effects of LOCA [loss-of-coolant

accident] + SSE [safe shutdown earthquake] on the steam generator

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

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

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

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

due to either the LOCA rarefaction wave and/or SSE loadings. Then

the resulting pressure differential on the deformed tubes may cause

some of the tubes to collapse.

There are two issues associated with steam generator tube

collapse. First, the collapse of steam generator tubing reduces the

RCS [reactor coolant system] flow area through the tubes. The

reduction in flow area increases the resistance to flow of steam

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

Peak Clad Temperature (PCT). Second, there is a potential the

partial through-wall cracks in tubes could progress to through-wall

cracks during tube deformation or collapse or that short through-

wall indications would leak at significantly higher leak rates than

included in the leak rate assessments.

Consequently, a detailed leak-before-break analysis was

performed and it was concluded that the leak-before-break

methodology (as permitted by GDC 4) is applicable to the Farley Unit

1 [RCS] 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 RCS primary loops, the LOCA loads from the large branch

line breaks were analyzed at Farley Unit 1 and were found to be of

insufficient magnitude to result in steam generator tube collapse or

significant deformation.

Regardless of whether or not leak-before-break is applied to the

primary loop piping at Farley, any flow area reduction is expected

to be minimal (much less than 1 [percent]) and PCT margin is

available to account for this potential effect. Based on analyses

results, no tubes near wedge locations are expected to collapse or

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

increased over current expected levels. For all other steam

generator tubes, the possibility of secondary-to-primary leakage in

the event of a LOCA + SSE event is not significant. In actuality,

the amount of secondary-to-primary leakage in the event of a LOCA +

SSE is expected to be less than that currently allowed, i.e., 500

gpd per steam generator. Furthermore, secondary-to-primary in-

leakage for the same pressure differential since the cracks would

tend to tighten under a secondary-to-primary pressure differential.

Also the presence of the tube support plate is expected to reduce

the amount of in-leakage.

Addressing the R.G. 1.83 considerations, implementation of the

tube repair criteria is supplemented by 100 [percent] inspection

requirements at the tube support plate elevations having outside

diameter stress corrosion cracking indications, reduced operating

leak rate limits, eddy current inspection guidelines to provide

consistency in voltage normalization, and rotating pancake coil

inspection requirements for the larger indications left in service

to characterize the principal degradation mechanism as outside

diameter stress corrosion cracking.

As noted previously, implementation of the tube support plate

elevation repair criteria will decrease the number of tubes which

must be taken out of service with tube plugs or repaired. The

installation of steam generator tube plugs or tube sleeves would

reduce the RCS flow margin, thus implementation of the interim

repair criteria will maintain the margin of flow that would

otherwise be reduced through increased tube plugging or sleeving.

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

does not result in a significant reduction in margin with respect to

plant safety as defined in the Final Safety Analysis Report or any

bases of the plant Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: James H. Miller, III, Esq., Balch and

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

Alabama 35201

NRC Project Director: S. Singh Bajwa

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: December 10, 1993

Description of amendment request: The proposed changes would modify

the surveillance frequency of the auxiliary feedwater system pumps and

valves from monthly to quarterly. Various administrative changes are

being proposed such as 1) punctuation and grammar, 2) correction of

system or component names, and 3) capitalization of defined words.

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:

Specifically, operation of Surry Power Station in accordance

with the proposed Technical Specifications changes will not:

1. Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated.

Changing the surveillance test frequency of the Auxiliary

Feedwater System pumps and valves does not significantly affect the

probability of occurrence or consequences of any previously

evaluated accidents. The probability of an accident occurrence is

not increased in itself by the proposed changes in surveillance

testing of the Auxiliary Feedwater System pumps and valves.

[A]uxiliary feedwater pump testing is performed through a full-flow

test line, thereby not affecting normal plant operations. Changes to

the testing therefore do not affect the probability of an accident

occurrence. Redundant trains of the Auxiliary Feedwater System

remain available during surveillance testing, therefore, the

consequences of an accident are unchanged by the proposed changes in

the Auxiliary Feedwater System surveillance test frequencies.

Quarterly testing of the pumps and valves will continue to assure

that the Auxiliary Feedwater System is capable of performing its

intended functions for either unit if called upon. Consistent with

Generic Letter 93-05, ``Line-Item Technical Specifications

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation,'' the new testing frequency should reduce Auxiliary

Feedwater System unavailability resulting from failures and

equipment degradation during testing, thereby resulting in improved

system reliability. Invoking ASME Section XI as the acceptance

criteria for testing the Auxiliary Feedwater Pumps is an enhancement

to the acceptance criteria presently specified. Furthermore, the

operability requirements for the Auxiliary Feedwater System remain

unchanged. Therefore, the probability or consequences of any

previously analyzed accident are not increased by the proposed

changes in surveillance requirements for the Auxiliary Feedwater

System.

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

from any accident previously evaluated.

Changes in test frequency and acceptance criteria for the

Auxiliary Feedwater System pumps and valves do not involve any

physical modification of the plant or result in a change in a method

of operation. Quarterly testing of the pumps and valves during both

operation and shutdown will continue to assure that the Auxiliary

Feedwater System will be capable of performing its intended function

for either unit. Invoking ASME Section XI acceptance criteria for

testing the Auxiliary Feedwater System pumps is an enhancement to

the acceptance criteria presently specified. The operability

requirements for the Auxiliary Feedwater System remain unchanged.

Furthermore, new or different failure modes are not introduced by

these changes in surveillance requirements. Therefore, a new or

different type of accident is not created by these proposed changes

in surveillance requirements for the Auxiliary Feedwater System.

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

Changing the surveillance requirements of the Auxiliary

Feedwater System pumps and valves does not affect any safety limits

or limiting safety system settings. System operating parameters are

unaffected. This reduced pump and valve testing frequency should

reduce Auxiliary Feedwater System unavailability due to actual

testing, as well as failures and equipment degradation during

testing. Thus reduced testing results in an improved system

reliability. Quarterly testing of the pumps and valves during

operation and shutdown will continue to assure that the Auxiliary

Feedwater System will be capable of performing its intended

functions for either unit. Therefore, the reduction in surveillance

testing requirements for the Auxiliary Feedwater System pumps and

valves does not reduce 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 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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

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

Virginia 23219.

NRC Project Director: Herbert N. Berkow

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: September 17, 1993

Description of amendment request: The proposed amendment would

revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications

(TS) by incorporating technical and administrative changes to TS 4.5,

Emergency Core Cooling System and Containment Air Cooling System Tests;

TS 4.7, Main Steam Isolation Valves; and Table TS 4.1-3, Minimum

Frequencies for Equipment Tests. Changes are proposed for the safety

injection (SI) system automatic initiation test; the internal

containment spray system (ICS) flow blockage test; the SI, ICS and

residual heat removal pumps' periodic tests; the main steam isolation

valves' test; and the periodic control rod functional test.

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) TS 4.5.a.1.A

The 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.

The probability of an accident previously evaluated is not

increased by the TS change. The changes do not affect any structure,

system, or component that initiates an accident analyzed in the

Updated Safety Analysis Report (USAR). The probability of an

accident occurring is independent of the availability of emergency

core cooling components used to mitigate an accident.

The consequences of an accident previously evaluated will not be

increased by this TS change. Revising the TS wording to clarify that

the pumps may be operated during the periodic surveillance tests

does not decrease their availability and therefore does not decrease

their ability to mitigate the consequences of accidents previously

evaluated in the USAR.

Clarifying the TS wording 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 accident previously evaluated.

A new or different kind of accident from those previously

evaluated in the USAR will not be created by this TS change.

The automatic SI actuation is designed to respond to various

events analyzed in the USAR which take credit for SI in the event

mitigation. The test required by TS 4.5.a.1.A verifies that the

valves, pump circuit breakers, and automatic circuitry receive the

SI signal in the proper sequence. The procedural prerequisites for

the performance of this test ensure that an adequate flow path and

overpressure protection are available for the pumps during the test.

This proposed amendment does not alter the plant configuration,

operating setpoints, or overall plant performance. It simply

provides clarification that the pumps may start and operate in

conjunction with the automatic circuitry test; however, the pumps

are not required to start and operate for this test. Therefore, the

possibility or a new or different accident from any previously

evaluated is not created.

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

The intent of the TS, i.e., verify each component receives the

SI signal in the proper sequence, is not altered by this proposed

change. Clarifying that the pumps are allowed to operate during the

performance of this surveillance requirement does not adversely

affect the ability of the surveillance to demonstrate system

actuation. Therefore, this proposed change will not reduce the

margin of safety.

(b) TS 4.5.a.2.B

The 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.

This proposed change does not affect the probability of an

accident previously evaluated. The ICS system is designed to respond

to various events analyzed in the USAR which take credit for spray

for event mitigation. The probability of an accident occurring is

independent of the availability of containment air cooling systems

used to mitigate an accident.

This proposed change does not affect the consequences of an

accident previously evaluated. The proposed change allows the use of

surveillance methods that provide assurance of system capability to

perform required design functions in the event of an accident. The

configuration of containment air cooling components used to mitigate

the consequences of accidents previously evaluated is not being

changed.

Therefore, the proposed changes do not increase the probability

or consequences of accidents previously evaluated.

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

from any accident previously evaluated.

A new or different kind of accident from those previously

evaluated will not be created by this TS change. The proposed

changes do not alter the operation, function or modes of plant or

equipment operation. Allowing the use of equivalent surveillance

methods does not create the probability of a new or different kind

of accident.

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

This proposed change does not alter the intent of the

specification. The use of equivalent test methods to demonstrate

component functionality does not reduce the margin of safety. The

proposed change does remove a requirement to verify 100% (i.e. 168)

of the installed nozzles are unobstructed; however, as noted in the

safety evaluation, the original system was overdesigned by

approximately 5%. Verification of the number of required nozzles to

satisfy design assumptions is assured by the design requirements

stated in the USAR. Since the minimum design requirements stated in

the USAR are demonstrated by surveillance, the margin of safety is

not reduced. Consistent with Section 3.6.6A and associated basis

statements of NUREG 1431, ``Westinghouse Standard Technical

Specifications,'' Revision 0, performing surveillance at ten year

intervals assures the required number of spray nozzles are capable

of functioning in the event of an accident. Therefore, there is no

adverse effect on public health and safety.

(c) TS 4.5.b.1.B

The 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.

The proposed change will not increase the probability of

accidents previously evaluated. The probability of an accident

occurring is independent of the availability of the systems used to

mitigate an accident.

The proposed change will not increase the consequences of an

accident previously evaluated. The consequences of previously

evaluated accidents requiring SI, RHR, and containment spray for

event mitigation are independent of the surveillance testing of

these systems.

The SI, RHR, and ICS pumps are designed to operate in response

to various events analyzed in the USAR which initiate on a SI or

high-high containment pressure signal. The proposed change is

intended to allow the use of full flow quarterly testing for the ICS

pumps and retain the existing quarterly mini-flow requirements for

the SI and RHR pumps. Surveillance testing verifies proper operation

of the components and thereby ensures the consequences of the

accident are consistent with the evaluations of the USAR.

Therefore, the proposed change will not result in a significant

increase in the probability or consequences of an accident

previously evaluated.

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

from any accident previously evaluated.

A new or different kind of accident from those previously

evaluated will not be created by this TS change. The proposed

amendment does not alter the plant configuration, operating

setpoints, or overall plant performance.

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

The margin of safety will not be reduced by verifying the pump's

performance through full flow testing or miniflow testing. The

intent of the TS, i.e. verifying no degradation of the ICS pump is

satisfied by performing a full flow test on a quarterly basis.

(d) Administrative changes to TS Section 4.5

The 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, 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 the margin of safety.

The proposed changes are administrative in nature and do not

alter the intent or interpretation of the TS. Therefore, no

significant hazards exist.

(e) TS Section 4.7

The proposed changes were 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.

The probability of an accident previously evaluated will not be

increased by this TS change. The MSIVs are designed to limit the

cooldown rate of the RCS and prevent structural damage to the

containment, resulting from a SLB incident, by closing to isolate

the SGs. The probability of a SLB occurring is independent of MSIV

operability.

The consequences of an accident previously evaluated will not be

increased by this TS change. This TS change allows testing of the

MSIVs when operating temperature and pressure conditions exist that

are consistent with the conditions under which the acceptance

criteria were generated. The acceptable closure time for the MSIVs

is not being modified by this TS change, therefore the consequences

of accidents relying on the closure of the MSIVs are not increased.

An evaluation of the Updated Safety Analysis Report (USAR) SLB

analyses was performed. The existing core power and reactor coolant

system transient analyses assumed initial hot shutdown conditions

for all cases analyzed since this represents the most conservative

initial conditions for the accident. The analysis of the containment

pressure transient also assumed initial hot shutdown conditions (at

which time the steam pressure is highest and there is the greatest

inventory of water in the steam generator). The containment pressure

analysis also conservatively delayed the MSIV closures such that

steam flow from both steam generators existed for the first ten

seconds. It has been determined that the proposed change in mode for

surveillance testing is enveloped by the existing analyses,

therefore the consequences of an accident remained unchanged.

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

from any accident previously evaluated.

A new or different kind of accident from those previously

evaluated will not be created by this TS change. The proposed

amendment does not alter the plant configuration, operating

setpoints or overall plant performance.

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

The margin of safety will not be reduced by this TS change.

Changing the plant conditions under which this surveillance is

performed does not alter the acceptance criterion for closure time.

Changing the operating mode for the surveillance test is based on

engineering judgement which recognizes the importance of

demonstrating the capability of the components to perform in a steam

environment.

(f) Table TS 4.1-3

The 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.

This proposed change will not alter the intent of this

specification which is to ensure that the control rods are operable

and capable of performing their safety-related function during a

USAR analyzed event. Control rods fully inserted into the core are

already performing their safety-related function and therefore, by

definition, are operable. Revising the specification clarifies that

the control rods fully inserted into the core need not be tested.

Therefore, the probability or consequences of an accident previously

evaluated will not be increased.

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

from any accident previously evaluated.

The proposed amendment does not alter the plant configuration,

operating setpoints, or overall plant performance. Therefore, a new

or different kind of accident from those previously evaluated will

not be created by this TS change.

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

Rods which are fully inserted into the core are already

performing their safety function and therefore, by definition, are

already operable. Exempting these control rods from a partial

movement test will not reduce the margin of safety.

(g) Administrative changes to Table TS 4.1-3

The 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, 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 the margin of safety.

The proposed changes are administrative in nature and do not

alter the intent or interpretation of the TS. Therefore, no

significant hazards exist.

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: John N. Hannon

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

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: November 23, 1993

Description of amendment request: The proposed amendments would

change the operating conditions and limiting conditions for operation

for containment systems, and revise corresponding definitions and

tests. In addition, related bases would be updated.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendments will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes will add operating conditions and limiting

conditions for operation to Section 15.3.6, ``Containment System,''

of the Point Beach Nuclear Plant (PBNP) Technical Specifications

(TS). This change also proposes revisions to Sections 15.1,

``Definitions,'' and 15.4.4, ``Containment Tests,'' to support the

changes to Section 15.3.6. Additionally, the change will add

explanatory text to the bases for Section 15.3.6 to support the

revisions.

The proposed revisions will add more specific limiting

conditions for operation (LCOs) for containment isolation valves,

air locks, overall containment air leakage, and internal pressure.

The proposed LCOs state more clearly the requirements for

operability and the actions to be taken if the requirements are not

met. The proposed LCOs reflect the intent of the Westinghouse

Improved Standard Technical Specifications, NUREG-1431. None of the

existing LCOs are being removed.

As most of the existing LCOs in Section 15.3.6 do not specify

completion times for required actions, Section 15.3.0, ``General

Considerations,'' applies. Section 15.3.0 states that if an LCO does

not prescribe a time period, the affected unit shall be placed in

hot shutdown within three hours of discovering the situation. If the

conditions which prompted the shutdown cannot be corrected, the unit

shall be placed in cold shutdown within 48 hours of discovering the

situation. The proposed revisions to Section 15.3.6 specify

appropriate completion times for each LCO to eliminate the need to

default to the generic time requirements given in Section 15.3.0. If

the required actions and associated completion times are not met,

the plant must be brought to at least hot shutdown within 6 hours

and to cold shutdown within 36 hours. The slightly longer time to

hot shutdown is more than offset by the shorter time to cold

shutdown, thereby reducing the consequences of a release from

containment. These times are reasonable, based on operating

experience, to reach the required plant conditions from full power

conditions in an orderly manner without challenging plant systems.

There is no physical change to the facility, its systems, or its

operation. Therefore, there is no increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

analyzed.

The proposed changes will add operating conditions and limiting

conditions for operation to Section 15.3.6, ``Containment System,''

of the Point Beach Nuclear Plant (PBNP) Technical Specifications

(TS). This change also proposes revisions to Sections 15.1,

``Definitions,'' and 15.4.4, ``Containment Tests,'' to support the

changes to Section 15.3.6. Additionally, the change will add

explanatory text to the bases for Section 15.3.6 to support the

revisions.

The proposed revisions will add more specific LCOs for

containment isolation valves, air locks, overall containment air

leakage, and internal pressure. The proposed LCOs state more clearly

the requirements for operability and the actions to be taken if the

requirements are not met. The proposed LCOs reflect the intent of

the Westinghouse Improved Standard Technical Specifications, NUREG-

1431. None of the existing LCOs are being removed.

As most of the existing LCOs in Section 15.3.6 do not specify

completion times for required actions, Section 15.3.0, ``General

Considerations,'' applies. Section 15.3.0 states that if an LCO does

not prescribe a time period, the affected unit shall be placed in

hot shutdown within three hours of discovering the situation. If the

conditions which prompted the shutdown cannot be corrected, the unit

shall be placed in cold shutdown within 48 hours of discovering the

situation. The proposed revisions to Section 15.3.6 specify

appropriate completion times for each LCO to eliminate the need to

default to the generic time requirements given in Section 15.3.0. If

the required actions and associated completion times are not met,

the plant must be brought to at least hot shutdown within 6 hours

and to cold shutdown within 36 hours. The slightly longer time to

hot shutdown is more than offset by the shorter time to cold

shutdown, thereby reducing the consequences of a release from

containment. These times are reasonable, based on operating

experience, to reach the required plant conditions from full power

conditions in an orderly manner without challenging plant systems.

There is no physical change to the facility, its system, or its

operation. Thus, a new or different kind of accident cannot occur.

3. The proposed amendments will not involve a significant

reduction in the margin of safety.

The proposed changes will add operating conditions and limiting

conditions for operation to Section 15.3.6, ``Containment System,''

of the Point Beach Nuclear Plant (PBNP) Technical Specifications

(TS). This change also proposes revisions to Sections 15.1,

``Definitions,'' and 15.4.4, ``Containment Tests,'' to support the

changes to Section 15.3.6. Additionally, the change will add

explanatory text to the bases for Section 15.3.6 to support the

revisions.

The proposed revisions will add more specific LCOs for

containment isolation valves, air locks, overall containment air

leakage, and internal pressure. The proposed LCOs state more clearly

the requirements for operability and the actions to be taken if the

requirements are not met. The proposed LCOs reflect the intent of

the Westinghouse Improved Standard Technical Specifications, NUREG-

1431. None of the existing LCOs are being removed.

As most of the existing LCOs in Section 15.3.6 do not specify

completion times for required actions, Section 15.3.0, ``General

Considerations,'' applies. Section 15.3.0 states that if an LCO does

not prescribe a time period, the affected unit shall be placed in

hot shutdown within three hours of discovering the situation. If the

conditions which prompted the shutdown cannot be corrected, the unit

shall be placed in cold shutdown within 48 hours of discovering the

situation. The proposed revisions to Section 15.3.6 specify

appropriate completion times for each LCO to eliminate the need to

default to the generic time requirements given in Section 15.3.0. If

the required actions and associated completion times are not met,

the plant must be brought to at least hot shutdown within 6 hours

and to cold shutdown within 36 hours. The slightly longer time to

hot shutdown is more than offset by the shorter time to cold

shutdown, thereby reducing the consequences of a release from

containment. These times are reasonable, based on operating

experience, to reach the required plant conditions in an orderly

manner without challenging plant systems.

There is no physical change to the facility, its systems, or its

operation. Thus, a significant reduction in a margin of safety

cannot occur. In fact, by adding more specific LCOs for containment

integrity and allowing for a more orderly unit shutdown when

required, an increased margin of safety may be realized.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: John N. Hannon

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 20555, and at the local public document

rooms for the particular facilities involved.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of application for amendment: October 22, 1993, as revised

October 28, 1993.

Brief description of amendment: The amendment (1) removes the

title-specific organizational listing of the Plant Nuclear Safety

Committee (PNSC) membership in TS 6.5.2.2 and replaces it with a

functional description of the PNSC composition, (2) adds specific

member qualification requirements in TS 6.5.2.3, and (3) revises

Section 6.5.2.2 to stipulate that PNSC members shall be designated by

the plant general manager and shall be limited in number to between

seven and nine members.

Date of issuance: December 28, 1993

Effective date: December 28, 1993

Amendment No.: 41

Facility Operating License No. NPF-63. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: November 24, 1993 (58

FR 62152) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 28, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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; Docket Nos. 50-373 and 50-374, LaSalle

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

Nos. 50-295 and 50-304, Zion Nuclear Power Station Units 1 and 2,

Lake County, Illinois

Date of application for amendments: November 10, 1993

Brief description of amendments: The amendments revised the

``Radioactive Effluent Controls Program'' described in Section 6.0 of

the Braidwood, Byron, LaSalle, and Zion Technical Specifications (TS)

to be consistent with the revised 10 CFR Part 20. The changes

specifically address the limitation on radioactive material release of

liquid and gaseous effluents.

Date of initial notice in Federal Register: November 24, 1993 (58

FR 62152) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 30, 1993, and an

environmental assessment noticed in the Federal Register December 30,

1993 (58 FR 69412).

Date of issuance: December 30, 1993

Effective date: December 30, 1993

Amendment Nos.: For Byron, 57 and 57; for Braidwood, 45 and 45; for

LaSalle, 93 and 77; for Zion, 152 and 140.

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

NPF-11, NPF-18, DPR-39, and DPR-48. The amendments revise the Technical

Specifications. No significant hazards consideration comments received:

No

Local Public Document Room locations: 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; for LaSalle, the Public Library of

Illinois Valley Community College, Rural Route No. 1, Oglesby, Illinois

61348; for Zion, the Waukegan Public Library, 128 N. County Street,

Waukegan, Illinois 60085.

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

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

Date of application for amendments: June 2, 1993

Brief description of amendments: The amendments would include

Commonwealth Edison Company's Topical Report NFSR-0091, in Section 6.6

of the Technical Specifications. Topical Report NFSR-0091 has been

reviewed and approved by the NRC staff.

Date of issuance: December 28, 1993

Effective date: Immediately, to be implemented within 45 days.

Amendment Nos.: 124 and 118

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57847). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 28, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Morris Public Library, 604

Liberty Street, Morris, Illinois 60450.

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

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

Date of application for amendment: May 5, 1993, as revised by

letter dated August 27, 1993, and supplemented by letter dated December

21, 1993.

Brief description of amendment: The amendment revises surveillance

intervals for the Containment Pressure Channels, the Steam Pressure

Channels, and the Reactor Coolant Temperature Channels to accommodate a

24-month refueling cycle. These revisions are being made in accordance

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

Technical Specification Surveillance Intervals to Accommodate a 24-

Month Fuel Cycle.''

Date of issuance: December 28, 1993

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

30 days.

Amendment No.: 167

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

Technical Specifications.

Date of initial notice in Federal Register: October 13, 1993 (58 FR

52981) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 28, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

Michigan

Date of application for amendment: March 23, 1993

Brief description of amendment: The amendment revises the

operability requirements and the surveillance frequency for periodic

testing of suppression chamber (torus) to drywell vacuum breakers. All,

vice 10 of 12, vacuum breakers are required to be operable and periodic

testing will be done each cold shutdown (if not performed within the

last 92 days) vice monthly.

Date of issuance: January 4, 1994

Effective date: January 4, 1994, with full implementation within 45

days.

Amendment No.: 96

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

Technical Specifications.

Date of initial notice in Federal Register: May 26, 1993 (58 FR

30191) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 4, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

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

Unit No. 2, Pope County, Arkansas

Date of application for amendment: July 22,1993

Brief description of amendment: The amendment relocated the

containment isolation valve table from the containment systems

specification to plant procedures, in accordance with Generic Letter

91-08.

Date of issuance: December 22, 1993

Effective date: December 22, 1993

Amendment No.: 154

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

Technical Specifications.

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46231) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 22, 1993. No

significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Mississippi Power &

Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit

1, Claiborne County, Mississippi

Date of application for amendment: September 23, 1993

Brief description of amendment: The amendment deleted the

surveillance requirement for jet pump differential pressure measurement

for operating loops when power is less than or equal to 25% rated

thermal power (RTP) and replaces the requirement with a provision

which, during operation at more than 25% RTP, allows continued

operation if the jet pump surveillance is performed within 24 hours

after RTP exceeds 25% or within 4 hours after an associated

recirculation loop is placed in service.

Date of issuance: January 4, 1994

Effective date: January 4, 1994

Amendment No: 110

Facility Operating License No. NPF-29. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: October 13, 1993 (58 FR

52984) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 4, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

Library, Post Office Box 1406, S. Commerce at Washington, Natchez,

Mississippi 39120.

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island

Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

Date of application for amendment: August 16, 1988, as supplemented

September 19, 1988; February 9, March 31, June 26, October 10, and

November 22, 1989; June 21, October 15, and November 7, 1990; February

19, April 19, June 21, August 28, and October 9, 1991; January 13,

1992; January 18, May 28, October 24, and November 12, 1993.

Brief description of amendment: This amendment replaces the TMI-2

Appendix A and B Technical Specifications with the Post-Defueling

Monitored Storage (PDMS) Technical Specifications to facilitate long

term storage of the facility.

Date of issuance: December 28, 1993

Effective date: December 28, 1993

Amendment No.: 48

Facility Operating License No. DPR-73: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 25, 1991 (56 FR

19128). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 28, 1993. No

significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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 application for amendment: October 21, 1993

Brief description of amendments: The amendments change the

technical specifications by revising Technical Specification 6.5.2.8,

``Audits,'' by removing the prescriptive frequency of the required

audits. The frequency has been removed in order to allow performance-

based inspection frequencies. This is consistent with NUREG-1431,

``Standard Technical Specifications - Westinghouse Plants,'' September

1992.

Date of issuance: December 27, 1993

Effective date: December 27, 1993, to be implemented within 10 days

of issuance.

Amendment Nos.: Unit 1 - Amendment No. 56; Unit 2 - Amendment No.

45

Facility Operating License Nos. NPF-76 and NPF-80. Amendment

revised the Technical Specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59752). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 27, 1993. 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

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: September 15, 1993, as supplemented by

letter dated November 30, 1993.

Brief description of amendments: The amendments consist of changes

to the technical specifications to implement the new requirements of 10

CFR Part 20 which were issued on May 21, 1991. The changes are either

editorial changes that provide consistency between the technical

specifications and the revised 10 CFR Part 20 or are changes to the

effluent limits cross referenced to the former 10 CFR Part 50, Appendix

I limits.

Date of issuance: December 30, 1993

Effective date: December 30, 1993, to be implemented on January 1,

1994.

Amendment Nos.: Unit 1 - Amendment No. 57; Unit 2 - Amendment No.

46

Facility Operating License Nos. NPF-76 and NPF-80: Amendment

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57854) The November 30, 1993, submittal provided additional clarifying

information and did not change the original proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

December 30, 1993, and an environmental assessment noticed in the

Federal Register December 27, 1993 (58 FR 68445). 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

Iowa Electric Light and Power Company, Docket No. 50-331, Duane

Arnold Energy, Center, Linn County, Iowa

Date of application for amendment: July 28, 1993

Brief description of amendment: The amendment revised the Technical

Specifications by incorporating the new requirements of 10 CFR Part 20.

The revision includes changing the release rate limits for gaseous and

liquid effluents, the monitoring and reporting requirements, the

definitions and the record retention requirements.

Date of issuance: December 27, 1993

Effective date: December 27, 1993

Amendment No.: 194

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

Technical Specifications.

Date of initial notice in Federal Register: September 15, 1993 (58

FR 48384) The Commission's related evaluation of the amendment is

contained in an Environmental Assessment dated December 23, 1993 (58 FR

68179), and a Safety Evaluation dated December 27, 1993. No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

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

Date of application for amendment: September 10, 1993, as

supplemented November 12, 1993

Brief description of amendment: The amendment changes the Technical

Specifications to provide consistency with the guidance of Generic

Letter 90-09 that relates to the revision of the surveillance

requirements for snubbers.

Date of issuance: December 28, 1993

Effective date: December 28, 1993

Amendment No.: 71

Facility Operating License No. DPR-21. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 29, 1993 (58

FR 50969). The November 12, 1993, letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

December 28, 1993. No significant hazards consideration comments

received: No.

Local Public Document Room location: Learning Resources Center,

Thames Valley State Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360.

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

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

Obispo County, California

Date of application for amendments: July 7, 1993 (Reference LAR 93-

05)

Brief description of amendments: The amendments revise the combined

Technical Specifications (TS) for the Diablo Canyon Power Plant Unit

Nos. 1 and 2 by changing the gaseous effluent limit of TS 6.8.4.g.,

``Radioactive Effluent Controls Program,'' and the Bases for TS 3/

4.11.1.4, ``Liquid Holdup Tanks,'' to conform to recent revisions to 10

CFR 20.

Date of issuance: January 6, 1994

Effective date: January 6, 1994

Amendment Nos.: 84, 85

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

revised the Technical Specifications.

Date of

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