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 March 31, 1995, through April 14, 1995. The

last biweekly notice was published on Wednesday, April 12, 1995 (60 FR

18621).

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 [[Page 20514]] margin of

safety. The basis for this proposed determination for each amendment

request is shown below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

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

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

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

intervene is discussed below.

By May 26, 1995, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, Attention: Docketing and Services

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee. [[Page 20515]]

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Carolina

Date of amendment request: March 20, 1995.

Description of amendment request: The licensee proposes a revision

to Technical Specification (TS) 2.2.1, Reactor Trip System

Instrumentation Setpoints, and to relocate cycle specific Overpower and

Overtemperature Delta T trip setpoint parameters to the Core Operating

Limits Report (COLR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

This change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change of relocating Overtemperature Delta T * * *

and the Overpower Delta T * * * trip setpoint parameters to the COLR

has no influence or impact to the probability or consequences of an

accident. The revised TS will continue to implement the Reactor Trip

System Instrumentation [Overtemperature Delta T] and [Overpower

Delta T] setpoint limits through reference to the parameters in the

COLR. In addition, the COLR is subject to the existing controls of

TS 6.9.1.6, including the establishment of the parameter values

using an NRC approved methodology. Given that this change

administratively relocates the selected trip setpoint parameter

values to another TS-controlled document, there would be no increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

No safety-related equipment, safety function, or plant operation

will be altered as a result of this proposed change. The limits are

simply being relocated to another TS-controlled document. The TS

will continue to require operation within the required limits as

established per NRC approved methodologies. Therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

Relocation of the Reactor Trip System Instrumentation

[Overtemperature Delta T] and [Overpower Delta T] setpoint limits to

the TS-controlled COLR has no effect on the trip system setpoints

currently in force in TS 2.2.1. Future revisions to the trip

setpoint parameters are governed by TS 6.9.1.6. TS 6.9.1.6 lists

each TS that references values in the COLR and the NRC approved

methodologies utilized in developing those values. Since this change

is only an administrative relocation of the selected trip setpoint

parameter values to another TS controlled document, 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 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: Cameron Village Regional

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

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

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

NRC Project Director: David B. Matthews.

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 amendment request: March 30, 1995.

Description of amendment request: The licensee proposes to revise

the Emergency Diesel Generator (EDG) surveillance requirements

contained in Technical Specification (TS) 4.8.1.1.2 to be consistent

with NUREG-1431, Standard Technical Specifications for Westinghouse

Plants, and to eliminate the need for duplicate EDG testing that has

already been implemented to satisfy the requirements of the Station

Blackout Rule and the Maintenance Rule.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

This change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

A failure of the Emergency Diesel Generator (EDG) is not an

initiator for any previously evaluated FSAR Chapter 15 accident

scenario. By committing to and implementing an EDG reliability

program that satisfies the requirements of the Station Blackout Rule

and the Maintenance Rule, the Shearon Harris Nuclear Power Plant

(SHNPP) will continue to ensure that target EDG reliability and

availability is being achieved by conducting appropriate monitoring,

testing, and maintenance activities. This program will be developed

and controlled as a Plant Operating Manual procedure and will

incorporate industry, vendor, and TDI Owners Group recommendations.

Therefore, with commensurate levels of testing and inspection in

place to provide assurance that the EDGs will perform their intended

safety function in the event of an accident, the proposed changes

will have no effect on the probability or consequences of such an

accident.

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

new or different kind of accident from any accident previously

evaluated.

A failure of the EDG is not an initiator for any previously

evaluated FSAR Chapter 15 accident scenario nor would the proposed

changes to the EDG surveillance requirements result in the

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

previously evaluated. By committing to and implementing an EDG

reliability program that satisfies the requirements of the Station

Blackout Rule and the Maintenance Rule, SHNPP will continue to

ensure that target EDG reliability and availability is being

achieved by conducting appropriate monitoring, testing, and

maintenance activities. This program will be developed and

controlled as a Plant Operating Manual procedure and will

incorporate industry, vendor, and TransAmerica Delaval Inc. Owners

Group recommendations. Therefore, with commensurate levels of

testing and inspection in place to provide assurance that the EDGs

will perform their intended safety function in the event of an

accident, the proposed changes would not increase the possibility of

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

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed changes will not affect any parameters which relate

to the margin of safety as defined in the Technical Specifications

or the FSAR. Testing, inspection and maintenance necessary to verify

the EDGs' ability to perform their intended safety function will

continue to be [[Page 20516]] performed. 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 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: Cameron Village Regional

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

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

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

NRC Project Director: David B. Matthews.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: March 24, 1995.

Description of amendment request: The proposed amendments would

acknowledge the acceptability of performing containment leakage rate

testing in accordance with 10 CFR Part 50, Appendix J, and all approved

exemptions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability of occurrence or consequences of any

accident previously evaluated.

The proposed changes to Technical Specifications to add an

allowance to test in accordance with approved exemptions to 10 CFR

50 Appendix J are administrative in nature and will not affect any

accident initiators or precursors. 10 CFR 50 Appendix J provides the

requirements to periodically test the primary reactor containment.

The objective of these requirements is to ensure that leakage from

the primary reactor containment structure and systems and components

that penetrate the containment is maintained below the limits

established for containment leakage. The performance of periodic

integrated leakage rate testing (Type A) and local penetration

testing (Type B and C) during containment life provides a current

assessment of potential leakage from containment during accident

conditions.

10 CFR 50.12 allows the Commission to grant specific exemptions

to the requirements of 10 CFR 50 Appendix J when those exemptions

are authorized by law, will not present undue risk to the public,

and are consistent with the common defense and security. In

addition, special circumstances must exist as described in Section

50.12. Since all exemptions to 10 CFR 50 Appendix J receive NRC

review and approval prior to being implemented, all containment

leakage rate testing will continue to be performed in accordance

with NRC approved methodologies when relying upon the allowance that

is added to the Technical Specifications by the proposed amendment.

The proposed changes are consistent with the requirements provided

in NUREG-1431, ``Standardized Technical Specifications, Westinghouse

Plants'' which has been approved by the NRC.

The proposed changes will not affect any accident initiators or

precursors and will not change or alter the design assumptions for

the systems used to mitigate the consequences of an accident. The

proposed changes do not involve the addition of any new or different

type of equipment, nor do they involve the operation of equipment

required for safe operation of the facility in a manner different

from those addressed in the UFSAR. There are no changes to

parameters governing plant operation as a result of the proposed

changes. The results and conclusions in the Zion Updated Final

Safety Analysis Report (UFSAR) are unaffected by this proposed

License Amendment.

Based on the previous discussion, the proposed changes do not

involve a significant increase in the probability of occurrence or

consequences of any accident previously evaluated.

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

new or different kind of accident from any previously analyzed.

The proposed changes to Technical Specifications to add an

allowance to perform containment leakage rate testing in accordance

with approved exemptions to 10 CFR 50 Appendix J are administrative

in nature and do not involve the addition of any new or different

types of safety related equipment, nor does it involve the operation

of equipment required for safe operation of the facility in a manner

different from those addressed in the safety analyses. The proposed

changes may only affect the methods used to perform containment

leakage rate testing while in a shutdown condition. No safety

related equipment or function will be altered as a result of the

proposed changes. Also, the procedures governing normal plant

operation and recovery from an accident are not changed by the

proposed Technical Specification changes. Since no new failure modes

or mechanisms are added by the proposed changes, the possibility of

a new or different kind of accident is not created.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

Given the proposed changes to Technical Specifications, testing

would be allowed in accordance with approved exemptions to Appendix

J. Exemptions are allowed by the Commission in accordance with 10

CFR 50.12 when it is shown that the exemption is authorized by law,

will not present an undue risk to the public health and safety, and

is consistent with the common defense and security. In addition,

special circumstances must exist.

The proposed changes will not impact any margin of safety and

testing in accordance with approved exemptions will not involve a

significant reduction in a level of safety since containment leakage

testing is performed while in a shutdown condition. In addition, it

is likely that any test methodology that significantly reduces a

margin of safety would not be approved by the NRC.

The ability to safely shut down the operating unit and mitigate

the consequences of all accidents previously evaluated will be

maintained. Therefore, the margin of safety is not significantly

affected.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

Austin, One First National Plaza, Chicago, Illinois 60690.

NRC Project Director: Robert A. Capra.

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

Charlevoix County, Michigan

Date of amendment request: December 15, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification 11.3.1.5 ACTION a. to eliminate the need

to demonstrate that the actuation circuitry of the unaffected reactor

depressurization system channels is operable. In addition, an editorial

change correcting a typographical error is also proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change will eliminate the probability of a

subsystem failure caused by additional testing (which unnecessarily

introduces the potential for human and equipment problems),

therefore eliminating the probability that the facility would have

to be challenged and brought to the SHUTDOWN condition within 12

hours and to the COLD SHUTDOWN condition within the following 24

hours.

[[Page 20517]]

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

different kind of accident from any accident previously evaluated?

The proposed change does not alter the plant configuration,

systems, components, or operation; and does not create the

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

previously evaluated. The proposed change is expected to eliminate

unnecessary challenges to a safety system that has already been

determined to be operable by routine surveillance testing; therefore

contributing to the overall safe operation of the facility.

3. Will the proposed change involve a significant reduction in

the margin of safety?

The RDS [Reactor Depressurization System] provides for both

manual and automatic depressurization of the primary system to allow

injection of the core spray following a small-to-intermediate size

break in the primary system. This will allow core cooling with the

objective of preventing excessive fuel clad temperatures. The design

of the system is based on the specified initiation set points

described in the Technical Specifications. Transient analysis

demonstrated that these conditions result in adequate safety margins

for both the fuel and the system pressure. The proposed change does

not affect these setpoints, therefore the margin of safety is not

changed.

In addition, the proposed editorial change to correct a

typographical error is administrative in nature and, therefore, would

have no effect on the three standards of 10 CFR 50.92 discussed above.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770.

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

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

NRC Project Director: Cynthia A. Carpenter, Acting.

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

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: January 26, 1995, as supplemented March

9, 1995.

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to increase the allowable

nominal fuel enrichment from 4.2 to 5.0 weight percent for reload fuel

assemblies. TS impose a limit on fuel enrichment of stored fuel

assemblies to prevent inadvertent criticality. Presently, the Crystal

River Unit 3 (CR3) TS specify a maximum enrichment of 4.5 weight

percent for storage pool A and dry fuel (new fuel) storage racks, and

4.2 weight percent for fuel pool B. The licensee proposed to revise TS

3.7.15, 4.2, and 4.3, and associated TS bases to allow increasing the

enrichment limits from 4.2 to 5.0 weight percent for the dry fuel

storage racks and for A and B fuel pools. Additionally, a typographical

error in TS 4.3.1.2.b will also be corrected.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. This amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

An increase in fuel enrichment will not by itself affect the

mixture of fission product nuclides. A change in fuel cycle design

which makes use of an increased enrichment may result in fuel burnup

consisting of a somewhat different mixture of nuclides. The effect

in this instance is insignificant because:

a. The isotopic mixture of the irradiated assembly is relatively

insensitive to the assembly's initial enrichment.

b. Most accident doses are such a small fraction of 10 CFR 100

limits, a large margin exists before any change becomes significant.

c. The change in Pu content which would result from an increase

in burnup would produce more of some fission product nuclides and

less of other nuclides. Small increases in some doses are offset by

reductions in other doses. The radiological consequences of

accidents are not significantly changed.

2. This amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

An unplanned criticality event will not occur as Keff

[effective neutron multiplication factor] will not exceed 0.95 with

the maximum allowable enriched fuel in Pool A and Pool B, when

flooded with unborated water, and Keff will not exceed 0.98 in

the new fuel storage racks assuming conditions of optimum

hypothetical low density moderation. The new fuel storage racks have

two rows of storage cells physically blocked to ensure reactivity

limits are not exceeded. Administrative controls assure fuel is

stored in configurations which meet the requirements of the safety

analysis.

3. This amendment will not involve a significant reduction in a

margin of safety.

While the increased enrichment in Pool A, Pool B, and the dry

storage racks may lessen the margin to criticality, this reduction

is not significant because the overall safety margin is within NRC

criteria of Keff [less than or equal to] 0.95 (NRC Standard

Review Plan, Section 9.1.2.)

Therefore, this amendment request satisfies the criteria

specified in 10 CFR 50.92 for amendments which do not involve a

significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629.

Attorney for licensee: A. H. Stephens, General Counsel, Florida

Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida

33733.

NRC Project Director: David B. Matthews.

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: March 16, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification 4.6.1.2, regarding the overall

integrated containment leakage rate tests, so that it would reference

10 CFR Part 50, Appendix J directly, rather than paraphrase the

regulation, and allow approved exemptions to the test frequency

requirements. In addition, there is an associated proposed exemption,

from the requirements of 10 CFR Part 50, Appendix J, to provide a one-

time interval extension for the Unit 2 Type A test (containment

integrated leak rate test) from the current scheduled 48 months to

approximately 66 months.

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:

Criterion 1--Does Not Involve a Significant Increase in the Probability

or Consequences of an Accident Previously Evaluated

The proposed change specific to Unit 2 will provide a onetime

exemption from the 10 CFR 50, Appendix J Section III.D.I.(a) leak

rate test schedule requirement. This change will allow for a one-

time test interval for Type A Integrated Leak Rate Tests of

approximately 66 months.

Leak rate testing is not an initiating event in any accident;

therefore, this proposed [[Page 20518]] change does not involve a

significant increase in the probability of a previously evaluated

accident.

Type A tests are capable of detecting both local leak paths and

gross containment failure paths. Experience at South Texas Project

Unit 2 demonstrates that excessive containment leakage paths are

local leakage detected by Type B and C Local Leak Rate Tests.

Administrative controls govern maintenance and testing of

containment penetrations such that the probability of excessive

penetration leakage due to improper maintenance or valve

misalignment is very low. Following maintenance on any containment

penetration, a Local Leak Rate Test is performed to ensure

acceptable leakage levels. Following a Local Leak Rate Test on a

containment isolation valve, an independent valve alignment check is

performed. Therefore, Type A testing is not necessary to ensure

acceptable leakage rates through containment penetrations.

While Type A testing is not necessary to ensure acceptable

leakage rates through containment penetrations, Type A testing is

necessary to demonstrate that there are no gross containment

failures. Structural failure of the containment is considered to be

a very unlikely event, and in fact, since South Texas Project Unit 2

has been in operation, it has successfully passed each Type A

Integrated Leak Rate Test. Therefore, a one-time exemption

increasing the interval for performing an Integrated Leak Rate Test

results [sic] in a significant decrease in the confidence in the

leak tightness of the containment structure. Therefore, this change

does not involve a significant increase in the consequences of an

accident previously evaluated.

The proposed amendment revised Technical Specification 4.6.1.2

to reference the testing frequency requirements of 10 CFR 50,

Appendix J, and to state that Nuclear Regulatory Commission approved

exemptions to the applicable regulatory requirements are permitted.

This portion of the proposed change is applicable to Unit 1 and Unit

2. The current language of Technical Specification 4.6.1.2

paraphrases the requirements of Section III,D.I.(a) [sic] of

Appendix J. The proposed administrative revision simply deletes the

paraphrased language and directly references Appendix J. No new

requirements are added, nor are any existing requirements deleted.

Any specific changes to the requirements of Section III.D.I.(a) will

require a submittal from Houston Lighting & Power under 10 CFR 50.12

and subsequent review and approval by the Nuclear Regulatory

Commission prior to implementation.

The proposed amendment, in itself, does not affect reactor

operations or accident analysis and has no radiological

consequences. The change provides clarification so that future

Technical Specification changes will not be necessary to correspond

to applicable Nuclear Regulatory Commission-approved exemptions from

the requirements of Appendix J.

Therefore, this proposed amendment does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind

of Accident From Any Previously Evaluated

The proposed Unit 2 exemption request does not affect normal

plant operations or configuration, nor does it affect leak rate test

methods. The proposed change allows a one-time test interval of

approximately 66 months for the Integrated Leak Rate Test. Because

the test history of South Texas Project Unit 2 demonstrates no Type

A test failures during plant lifetime, the relaxation in schedule

should not significantly decrease the confidence in the leak

tightness of the containment.

The proposed Technical Specification amendment for Units 1 and 2

provides clarification to a specification that paraphrases a

codified requirement.

Since the proposed change and amendment would not change the

design, configuration or method of operation of the plant, they

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

accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of

Safety

The purpose of the existing schedule for Integrated Leak Rate

Tests is to ensure that release of radioactive materials will be

restricted to those leak paths and leak rates assumed in accident

analyses. The relaxed schedule for Integrated Leak Rate Tests does

not allow for relaxation of Type B and C Local Leak Rate Tests.

Therefore, methods for detecting local containment leak paths and

leak rates are unaffected by this proposed change. A one-time

increase of the South Texas Project Unit 2 test interval does not

leak to a significant probability of creating a new leakage path or

increased leakage rates because the test history for Integrated Leak

Rate Tests shows no failure during plant life. The margin of safety

inherent in existing accident analyses is maintained.

The proposed Technical Specification amendment for Units 1 and 2

is administrative and clarifies the relationship between the

requirements of Technical Specification 4.6.1.2, Appendix J, and any

approved exemptions to Appendix J. It does not, in itself, change a

safety limit, a Limiting Condition of Operation, or a surveillance

requirement on equipment required to operate the plant. Nuclear

Regulatory Commission approval of any proposed change or exemption

to III.D.1.(a) of Appendix J will be required prior to

implementation.

Therefore, this change and amendment 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: William D. Beckner.

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

County, Iowa

Date of amendment request: March 10, 1995.

Description of amendment request: The proposed amendment would

remove redundant Limiting Conditions of Operation and Surveillance

Requirements for the containment hydrogen and oxygen monitors in the

Technical Specifications (TS).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. No physical changes will result from this

amendment. This change deletes requirements that are redundant and

unduly restrictive. The annual surveillance deleted by this

amendment is redundant to the semi-annual surveillance required in

Table 4.2-H. The Limiting Conditions for Operation are not changed

by the proposed amendment.

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

new or different kind of accident from any accident previously

evaluated. No physical changes will result from this amendment.

Functional tests are performed on the hydrogen and oxygen analyzers

semiannually as required in TS Table 4.2-H. Deleting the annual

requirement for a functional test of the same equipment will not

reduce the amount of testing performed or increase the possibility

of degraded equipment being undetected.

3. The proposed amendment does not involve a significant

reduction in a margin of safety. No physical changes will result

from this amendment. The existing requirement for a semi-annual test

of the hydrogen and oxygen analyzer in Table 4.2-H exceeds the

requirements to be deleted in Section 3.7/4.7-H. The frequency of

testing of the hydrogen and oxygen analyzers will not be reduced as

a result of this amendment.

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

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

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

amendment request involves no significant hazards consideration.

[[Page 20519]]

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis

& Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Gail H. Marcus.

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

County, Iowa

Date of amendment request: March 28, 1995.

Description of amendment request: The proposed amendment would

revise and clarify Technical Specification Table 3.2-A that lists

allowable out-of-service times and surveillance test intervals for

instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes to TS Table 3.2-A will not significantly

increase the probability or consequences of an accident previously

evaluated. The changes do not alter the physical design or operation

of the plant and serve to describe more accurately and clearly the

actual logic configurations. The existing logic designs are in

conformance with the Architect/Engineer's design documentation since

plant startup. These changes will assure that the information in the

tables is clearer and more consistent with the column headings of

the table. The proposed changes do not affect assumptions contained

in the plant safety analysis.

The Bases changes provide additional information about the logic

arrangements as appropriate to identify unique or different logic

configurations. Changes to the Allowed Outage Time (AOT)

descriptions for the MSL Flow--High and MSL Tunnel Temperature--High

provide clarification regarding application of the AOT to these

logic arrangements, since multiple instrument channels provide input

into multiple logic channels. This application conforms to the

single failure criterion of the design basis (NEDO-10139, Compliance

of Protection Systems to Industry Criteria: General Electric BWR

Nuclear Steam Supply System, dated June 1970) and to the analytical

basis for the TS (NEDC-31677P-A, Technical Specification Improvement

Analysis for BWR Isolation Actuation Instrumentation, dated July

1990).

2. The proposed changes to Table 3.2-A will not introduce a new

or different kind of accident from any accident previously

evaluated. The changes do not alter the physical design of the plant

or affect any modes of operation. The proposed changes serve to

clarify the existing information to better assure that the trip

instrumentation will be maintained as assumed in the accident

analyses contained in the Updated Final Safety Analysis Report.

3. The proposed changes do not involve a significant reduction

in a margin of safety. Clarification of the logic arrangements in

both Table 3.2-A and the TS Bases and how the AOT is applied does

not affect the ability of the isolation logic to perform its

intended function. No physical changes to the plant are being made

as part of this amendment.

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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis

& Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien County, Michigan

Date of amendment request: March 17, 1995.

Description of amendment request: The proposed amendment would

defer performance of the Type A containment integrated leakage rate

test until the next refueling outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Per 10 CFR 50.92, a proposed change does not involve a

significant hazards consideration if the change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated,

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

from any accident previously evaluated, or

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

Criterion 1

The Cook Nuclear Plant Type A test history provides substantial

justification for the proposed test schedule. Three Type A tests

were performed over a seven year period with successful results. The

tests indicate that the Cook Nuclear Plant has a low leakage

containment. In addition, there are no adverse trends in the results

from the previous Types A, B, and C tests or visual inspections that

indicate a gradual degradation of the containment boundary. Further,

there are no structural modifications planned which would adversely

affect the structural capability of the containment and that would

be a factor in deferring the Type A test one refueling outage.

Containment leak rate testing is not an initiator of any accident.

The proposed interval extension does not affect reactor operations

or the accident analysis and has no radiological consequences,

except for the dose savings associated with not performing the test.

There will be no changes to 10 CFR 100 dose limits or the control

room dose limits. Extending the test interval will not increase the

probability of a malfunction of equipment important to safety. Based

on these considerations, it is concluded that the change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Criterion 2

The proposed change does not involve physical changes to the

plant or changes in plant operating configuration. The proposed

change only relaxes the schedular requirements for conducting one

Type A test from the T/Ss and defers performance of the test one

cycle. The purpose of the test is to provide periodic verification

of the leak-tight integrity of the primary reactor containment, and

systems and components which penetrate containment. The tests assure

that leakage through containment and systems and components

penetrating containment will not exceed the allowable leak rate

values established in 10 CFR 50, Appendix J. Thus, it is concluded

that the proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Criterion 3

The proposed change to the schedule for performing the Type A

test does not reduce the margin of safety assumed in the accident

analysis for any release of radioactive materials or reduce any

margin of safety preserved by the technical specifications. The

methodology, acceptance criteria, and the technical specification

leak rate limits for the performance of the Type A test will not

change. Type A tests will continue to be performed in accordance

with 10 CFR 50, Appendix J and the applicable Cook Nuclear Plant

Technical Specifications beginning in 1997. In addition, there are

no adverse trends in the results from the previous Type A, B, and C

tests or visual inspections that indicate a gradual degradation of

the containment boundary. Based on these considerations, it is

concluded that the change does not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.

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

and [[Page 20520]] Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Project Director: Cynthia A. Carpenter, Acting.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

Date of amendment requests: March 31, 1995.

Description of amendment requests: The proposed amendments would

modify the Containment Ventilation System Technical Specifications (and

associated Bases) to allow limited containment purge operation in Modes

1, 2, 3, and 4 for pressure control, ALARA [as low as is reasonably

achievable], and respirable air quality considerations.

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:

Criterion 1

The purpose of this amendment is to allow flexibility in the use

of the containment purge system during MODES 1, 2, 3, and 4. The use

of this system during these modes of operation has previously been

approved (Amendment No. 66). Therefore, this amendment request does

not involve a significant increase in the probability or

consequences of an accident previously evaluated because the

proposed change to the T/Ss does not affect the assumptions,

parameters, or results of any UFSAR [Updated Final Safety Analysis

Report] accident analysis. Based on the existing system design and

demonstrated closure capability it is concluded that the proposed

changes do not modify the response of the containment during a

design basis accident. The proposed amendment does not add or modify

any existing equipment. Based on these considerations, it is

concluded that the changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Criterion 2

The proposed change does not involve physical changes to the

plant or changes in the plant operating configuration. Thus, it is

concluded that the proposed change does not create the possibility

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

evaluated.

Criterion 3

The margin for safety presently provided is not reduced by the

proposed change. As discussed previously, the containment purge

valves have been designed and demonstrated capable of closure

against the dynamic forces resulting from a loss of coolant

accident. The proposed amendment does not impact the ability of the

purge valves to perform their intended function (i.e. achieve

closure) in the event of an accident. Based on these considerations,

it is concluded that the 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 10 CFR 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.

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

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

NRC Project Director: Cynthia A. Carpenter, Acting.

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

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

Date of amendment request: March 31, 1995.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to increase the as-found

setpoint tolerance of the safety/relief valves (SRVs) from plus or

minus 1% to plus or minus 3%. In addition, the proposed amendment (1)

would allow the as-found condition of one SRV to be inoperable, (2)

clarifies the 1325 psig safety limit wording, (3) increases the number

of SRVs to be tested during each refueling outage, (4) makes editorial

changes to reflect the TS changes, and (5) revises the bases for the

applicable sections.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed changes in accordance with 10

CFR 50.92 and concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

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

compromised. The proposed changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The safety function of the SRVs is to mitigate the effects of a

RPV [reactor pressure vessel] overpressurization, therefore a

failure to open until the upper setpoint limit (+3%) is reached

cannot affect the probability of an accident. The lowest allowable

limit (-3%) is still above normal operating pressure and therefore

does not significantly increase the probability of an inadvertent

opening.

Should the SRVs open in response to an RCS [reactor coolant

system] overpressure event, opening of the SRVs below the nominal

setpoints does not adversely affect the consequences of an accident.

The fuel reload analysis demonstrates that actuation of five valves

at or below 103% of nominal provides sufficient pressure reduction

to maintain peak RCS pressure below the safety limit of 1375 psig

and to maintain vessel steam space pressure below 1325 psig. The

hydrodynamic loads on the SRV discharge pipe (i.e., tail pipe) and

the torus remain within the design limits.

The performance of the high pressure systems; FWCI [feedwater

coolant injection], SLC [standby liquid control] and IC [isolation

condenser] remain acceptable. There is also no adverse impact on the

operability of the APR [automatic pressure relief] system.

The SRV setpoints will continue to be required to be within

[plus or minus] 1% prior to plant startup from a refueling outage.

This ensures that the SRVs are restored to the optimal conditions at

the start of each fuel cycle.

Therefore, increasing the ``as-found'' tolerance from [plus or

minus] 1% to [plus or minus] 3% does not result in a significant

increase in the probability or consequences of a previously analyzed

accident.

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

from any previously analyzed.

Revising the acceptable as-found setpoint tolerance from [plus

or minus] 1% to [plus or minus] 3% does not change the type of

action that these valves are expected to perform, nor does it change

the initial ``as-left'' requirements for the valves. Plant operating

parameters have not changed. Therefore, this change cannot create

the possibility of a new or different kind of accident.

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

The margin of safety established and stated in the Millstone

Unit No. 1 Technical Specifications, is a peak RCS pressure of 1375

psig and a peak vessel steam space pressure of 1325 psig. While

allowing the SRV setpoint tolerance to increase to [plus or minus]

3% would allow peak pressures from an MSIV [main steam isolation

valve] closure event to approach that safety limit, the safety limit

will not be exceeded. Therefore, this change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360. [[Page 20521]]

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

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

06141-0270.

NRC Project Director: Phillip F. McKee.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Dockets

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: February 10, 1995.

Description of amendment request: The proposed changes provide for

the correction of administrative errors made in the past during the

processing of technical specification changes related to control room

ventilation filter surveillance testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated

because the changes are purely administrative and do not involve any

physical changes to plant SSC [systems, structures, or components].

Therefore, these changes will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated

because the changes will not alter the plant or the manner in which

the plant is operated. The changes do not allow plant operation in

any mode that is not already evaluated in the safety analysis. The

changes will not alter assumptions made in the safety analysis and

licensing bases. Therefore, these changes will not create the

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety because they are purely administrative and

have no impact on any safety analysis assumptions.

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, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

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

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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

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

Pennsylvania

Date of amendment request: March 15, 1994.

Description of amendment request: This amendment would reflect an

exemption from 10 CFR Part 50, Appendix J, Section II.H.4, concerning

the scope of Type `C' testing on specified emergency core cooling

system and reactor core isolation cooling containment isolation valves

by revising Technical Specification Table 3.6.3-1, Primary Containment

Isolation Valves. The subject valves on systems which terminate below

the minimum water level of the suppression pool and are associated with

closed systems would be tested using requirements of the American

Society of Mechanical Engineers' Section XI Code.

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:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to the scope of Type `C' testing for the

subject valves does not affect the probability of the design basis

accidents. The valves will continue to be maintained in an operable

state, and in their current design configuration. There is no

correlation between the scope of the Type `C' testing and accident

probability.

PP&L reviewed the postulated consequences of design basis events

on primary containment isolation under the proposed change. GDC 50

design conformance states that the primary containment structure,

including access openings, penetrations and the containment heat

removal system, is designed so that the containment structure and

its internal compartments can withstand, without exceeding the

design leakage rate (1.0% per day), the peak accident pressure and

temperature that could occur during any postulated LOCA.

For the purposes of considering the consequences of LOCAs under

the proposed change, a single active failure of a CIV or a passive

failure of the closed system were reviewed, within the limits of the

existing licensing basis. Under the existing licensing basis, a pipe

rupture of seismically qualified ECCS piping does not have to be

assumed concurrent with the LOCA, except if it is a consequence of

the LOCA. Consequential failures can be eliminated, since a LOCA

inside containment is separated from the ECCS piping by the

containment structure. Consequential failures of the ECCS piping

from LOCA's outside containment are outside the Appendix J design

considerations, although they are adequately addressed through the

redundancy and separation of the ECCS design. A single active

failure of the CIV, under the LOCA condition, can be accommodated

since the closed and filled system piping remains as the leakage

barrier. The ECCS passive failure criterion does require

consideration of system leaks, but not pipe breaks, beyond the

initiating LOCA. Pipe leakage, equivalent to the leakage from a

valve or pump seal failure, should be considered at 24 hours or

greater post-LOCA. The capability to make-up inventory to the

suppression pool is adequate to ensure that postulated seat leakage

and pipe leakage does not result in a condition that jeopardizes

pool level. Make-up capability exists to the suppression pool via

the Condensate Storage Tank and Spray Pond. Actions to make-up to

the suppression pool are delineated in Emergency Operating

Procedures.

Therefore, the proposal to eliminate the subject Type `C' tests

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The acceptability of the proposed change to the scope of Type

`C' testing for the subject valves is based on maintaining the

existing barriers to primary containment leakage, and ensuring that

the suppression pool level is assured for 30 days during all design

basis, post-accident modes of operation. By meeting these dual

objectives, the plant response to the design basis events will be

unchanged, and no new accident scenarios will be encountered. These

two objectives are related, in that, the suppression pool inventory

creates a passive barrier to primary containment atmospheric leakage

for penetrations which are located below the minimum water level of

the pool. The subject valve lines terminate below the minimum

suppression pool water level.

The subject valves are all single isolation valves associated

with lines that penetrate the primary containment, but are not

connected directly to the primary containment atmosphere or the

reactor coolant pressure boundary. The redundant isolation boundary

for each of the affected valves is the closed system associated with

the valve. This configuration is described in General Design

Criteria (GDC) 57. The proposed exemption, and Technical

Specification change, does not alter the configuration of these

systems. The valves will continue to be tested and maintained to

ensure their operability. The closed system [[Page 20522]] piping

meets PP&L's design conformance to GDC 56 and is verified via a

10CFR50 Appendix J Type `A' test. The integrity of the closed

systems is also monitored and controlled via Technical Specification

6.8.4.a.

The subject valves may be open, or change state, postaccident to

support the design function of their associated ECCS systems (HPCI,

Core Spray, RHR) or RCIC. The subject valves function as system

valves during the periods when they are open or in an intermediate

state, not as containment isolation valves. Reliance is placed on

the suppression pool seal and the closed system piping to maintain

the barrier between primary and secondary containment atmospheres.

Therefore, with the valve and closed system configuration

unaffected by the proposed change, the existing barriers to primary

containment atmospheric leakage are maintained, so long as the

suppression pool level is ensured.

The suppression pool is designed and operated so that it is

filled with water in accordance with Technical Specifications 3/

4.5.3, ``Suppression Chamber,'' 3/4.6.2, ``Depressurization

Systems--Suppression Chamber,'' and the associated Bases. The supply

of water in the suppression pool is assured for 30 days during all

design basis, post-accident modes of operation. Type `C' leak rate

testing has historically been performed on valves associated with

lines that connect to the suppression pool. The acceptance criteria

for combined leakage from these penetrations is 3.3 gpm. This

leakage rate is at a level which ensures the 30 day post-accident

suppression pool level. However, for the valves discussed in this

change, seat leakage past the CIV is into a closed and filled

system. Thus ``leakage'' from the suppression pool, past the CIV, is

a function of closed system leakage.

As mentioned above, the integrity of the closed system piping is

verified via a 10CFR50 Appendix J Type `A' test and is monitored and

controlled via Technical Specification 6.8.4.a. TS 6.8.4.a

establishes a program to monitor and control leakage from systems

located outside containment that could contain highly radioactive

fluids during a serious transient or accident. This program applies

to the ECCS systems and RCIC affected by the proposed change and

ensures that leakage into secondary containment via packing,

flanges, seals, etc., is controlled. Leakage from these systems,

plus the Scram Discharge Volume, Reactor Water Clean-up, and PASS,

has been found to be very low, and well below the 5 gpm limit

established for these systems. Current leakage for Unit 1 is 0.14

gpm and for Unit 2, 0.043 gpm. The proposed change is not expected

to contribute to higher levels of system leakage. Any leakage from

these systems is processed via Standby Gas Treatment and the

radwaste system to maintain ALARA and comply with regulatory

guidance. The closed systems are maintained filled, so that a supply

of water exists on both sides of the isolation valves.

While suppression pool leakage is a function of closed system

leakage for the subject penetrations, a review of Type `C' test data

for the subject CIVs showed that the valves have had low leakage

rates during previous tests. This leakage is on the order of 0.6

gpm, per unit. Proposed testing of the valves under Section XI and

the current requirements of the Generic Letter 8910 program will

ensure valve operability.

Therefore, leakage past the CIV and out of the closed system is

expected to be low and in keeping with the design basis for the

suppression pool. However, the capability does exist to make-up

water to the suppression pool from the Condensate Storage Tank or

Spray Pond if necessary. Existing Emergency Operating Procedures

require actions if suppression pool level is less than 22 feet or

greater than 24 feet. Thus, the level of the suppression pool is

ensured, independent of the current CIV Type `C' testing

requirement.

The proposed change to the scope of Type `C' testing for the

subject valves maintains the existing barriers to primary

containment leakage, and ensures that the suppression pool level is

assured for 30 days during all design basis, post-accident modes of

operation. Therefore, the plant response to the design basis events

is unchanged, and the proposal does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

As discussed in questions I and II, the proposed change does not

alter the plant response to existing accident scenarios, and does

not introduce new or different scenarios. So the margin of safety

from a design basis accident standpoint is maintained.

Historically, the leakage rate through the subject valves has

been determined under the Type `C' testing program. This leakage

rate has been found to be very low, and is currently on the order of

0.6 gpm. Quantifying leakage past the CIVs has been used to ensure

that the suppression pool level is assured for 30 days post-

accident. Under the proposed change, this leakage rate will not be

quantified. This is acceptable since leakage from the suppression

pool is in reality a function of closed system leakage, not solely

CIV leakage. Closed system leakage is monitored and controlled by an

existing Technical Specification program. Closed system leakage has

been found to be very low on both units, and is currently a small

fraction of a gallon per minute compared with a 5 gpm allowable.

Therefore, leakage past the CIV and out of the closed system is

expected to be low and in keeping with the design basis for the

suppression pool. However, the capability does exist, and is

proceduralized, to make-up water to the suppression pool from the

Condensate Storage Tank or Spray Pond if necessary. Thus the current

capability to maintain adequate suppression pool level for 30 days

postaccident is assured under the proposed change.

Therefore the proposed change to the scope of Type `C' testing

for the subject valves 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: Osterhout Free Library,

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

Pennsylvania 18701.

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

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

NRC Project Director: John F. Stolz.

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

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

Pennsylvania

Date of amendment request: March 31, 1995.

Description of amendment request: These amendments would modify the

surveillance requirement for reactor coolant system pH analysis in

section 4.4.4 of the Technical Specifications (TS) for each unit. Also,

they would clarify in the TS that the pH analysis would be taken at

least every 72 hours whenever reactor coolant conductivity exceeds 1.0

mho/cm.

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:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The pH limits on reactor coolant are not affected by this

change. The pH will be measured whenever it is theoretically

possible for it to be outside the Tech Spec [Technical

Specifications] limits of 8.6 (i.e., whenever the

conductivity is greater than 1.0 mho/cm). Because of the

theoretical relationship between pH and conductivity as shown in

Attachment A [see application dated March 31, 1995, for this

reference], it is possible to establish pH limits on the reactor

coolant by limiting the conductivity. As shown in this figure, the

pH must be >5.6 and mho/cm. Attachment A was taken from Regulatory Guide

1.56 Revision 1, July 1978 ``Maintenance of Water Purity in Boiling

Water Reactors''. As noted in both FSAR final safety analysis report

and Technical Specification Bases, the pH and conductivity limits

for OPERATIONAL CONDITION 1 are consistent with this theoretical

relationship. The Bases for Section 3/4.4.4 of the Tech Specs

[Technical Specifications] contains [contain] the following

statement: ``When the conductivity is within limits, the pH,

[[Page 20523]] chlorides and other impurities affecting conductivity

must also be within their acceptable limits[''].

Since the conductivity is measured by grab sampling at least

every 72 hours to verify that it is within limits, this will also

verify that pH is within limits every 72 hours. If the conductivity

should exceed 1.0 mho/cm, pH measurements will be made to

determine if the Tech Spec [Technical Specifications] pH limits have

been exceeded. It should also be noted that inline conductivity

instrumentation is very stable and reliable and is used to

continuously monitor the reactor coolant per Tech Spec [Technical

Specifications] requirements, with instrumentation connected to

redundant sources (reactor water cleanup influent and reactor

recirculation loop). Therefore, the proposed change will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

As stated above, the pH limits on reactor coolant are not

affected by this change. Since the conductivity is monitored

continuously, to verify that it is within limits, this will also

verify that pH is within limits. If the conductivity should exceed

1.0 mho/cm, pH measurements will be made to determine if

the Tech Spec [Technical Specifications] pH limits have been

exceeded. Therefore, the incorporation of this change will not

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

any accident previously evaluated.

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

The in-line conductivity instrumentation has been determined to

be very stable and reliable in its use to continuously monitor the

reactor coolant per Tech Spec [Technical Specifications]

requirements. To maintain this reliability, this instrumentation is

connected to redundant sources (reactor water cleanup influent and

reactor recirculation loop). Based on this continuous monitoring of

reactor coolant conductivity, as provided by this instrumentation,

the incorporation of this change will have no impact on current

safety margins, nor will it 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: Osterhout Free Library,

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

Pennsylvania 18701.

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

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

NRC Project Director: John F. Stolz.

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

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

Pennsylvania

Date of amendment request: March 31, 1995.

Description of amendment request: These amendments would delete

from the Technical Specifications of each unit, the operational

condition restriction in Surveillance Requirement 4.8.1.1.2.d.7 which

requires that 24-hour emergency diesel generator testing be performed

with at least one unit in operational condition 4 or 5 (cold shutdown

or refueling).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below: The proposed changes do not:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to permit the 24 hour testing of the

emergency diesel generators during power operation does not increase

the chances for a previously analyzed accident to occur. The

function of the EDGs [emergency diesel generators] is to supply

emergency power in the event of a loss of offsite power. As stated

above [,] the diesel generator being tested has been determined to

remain operable and available to supply the emergency loads within

the required times. In addition, the three remaining EDGs will be

operable during this test. Operations [Operation] of an EDGs [EDG]

is not a precursor to any accident. If, however, an offsite

disturbance were to occur that affected the operability of the DG

[emergency diesel generator] being tested, the remaining EDGs are

capable of feeding the loads necessary for safe shutdown of the

plant. In summary, the proposed change does not adversely affect the

performance or the ability of the diesel generators to perform their

intended safety function. Therefore, the proposed change will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

The proposed change to the 24 hour surveillance requirement will

not affect the operation of any safety system or alter its response

to any previously evaluated accident. The diesel generator will

automatically transfer from test mode if necessary to supply

emergency loads in the required time. The test mode is used for the

monthly surveillances of these diesel generators, resulting in no

new plant operating modes being introduced. In the event the EDG

fails the functional test[,] it will be declared inoperable and the

actions required for an inoperable diesel will be performed. The

remaining three EDGs will be operable and are capable of feeding the

loads necessary for safe shutdown of the plant. Therefore, the

incorporation of this change will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

Changing the EDG test timing results is no reduction in the

safety margin as defined in the design basis. Because loss of an EDG

is not expected as a result of LOOP [loss of offsite power] or LOCA/

LOOP [loss-of-coolant accident with a loss of offsite power] during

the 24 hour test, SSES [Susquehanna Steam Electric Station] remains

within its design basis. In fact, because the test EDG loads the ESS

[engineered safety system] bus 8.5 seconds earlier than the non-test

EDGs during LOCA with LOOP, plant response is actually improved.

Risk of operation during the 24 hour EDG test is certainly less than

during the current 84 hour allowed outage time (AOT) because both

the impact of the initiating events evaluated (EDG in test is not

actually failed) and the frequency of the limiting plant condition

(loss of two EDGs) are less. No increase in frequency or impact of

design basis events, and no reduction in the safety margin occurs

during the 24 hour EDG test. Therefore, the incorporation of this

change will have no impact on current safety margins, nor will it

involve a significant reduction in the margin to [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: Osterhout Free Library,

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

Pennsylvania 18701.

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

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

NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna

Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: March 31, 1995.

Description of amendment request: This amendment would change

Susquehanna Unit 2 Technical Specifications (TS) by incorporating the

General Electric (GE) NRC approved methodology for GE-12 type lead use

fuel assemblies (NEDE-24011-P-A-10) into the list of references in

Section 6.9.3.2. The licensee plans to insert four of these fuel

assemblies into the Unit 2 core during the fall of 1995. The addition

of the reference to the TS would allow the use of the GE methodology to

document that all [[Page 20524]] applicable requirements of the safety

analysis are met by the assemblies.

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 does not:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Incorporation of this proposed change of adding reference NEDE-

24011-P-A-10, ``General Electric Standard Application for Reactor

Fuel'' to the list of references in [the] Unit 2 Technical

Specifications will allow the use of the GE methodology to calculate

the operating limits for the four GE Lead Use Assemblies which are

of a different mechanical design than the Siemens 9X9 fuel

[currently installed in the reactor core]. This NRC approved

methodology will be referenced as the approved methodology in

showing that all applicable safety limits of the safety analysis are

met by the four GE-12 LUAs. Results of incorporating this change

will not significantly increase the probability or the consequences

of an accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

As stated above, the incorporation of this change will allow the

use of the GE methodology to be referenced as the approved

methodology to show that all applicable limits of the safety

analysis are met by the four GE-12 LUAs. Therefore, the

incorporation of this change will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

The use of the GE methodology will not result in a change in

safety margin, but will ensure that the safety margin is maintained

with the insertion of the four GE LUAs of the GE-12 type in Unit 2

Cycle 8. Therefore, the incorporation of these changes will have no

impact on current safety margins, nor will they involve a

significant reduction in the margin to [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: Osterhout Free Library,

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

Pennsylvania 18701.

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

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

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: December 2, 1994.

Description of amendment request: The proposed change to Limerick

Generating Station, Units 1 and 2, Technical Specifications (TS)

relocates the TS Fire Protection Requirements to Licensee controlled

documents consistent with NRC Generic Letter (GL) 86-10

``Implementation of Fire Protection Requirements,'' and GL 88-12,

``Removal of Fire Protection Requirements from Technical

Specifications.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications 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 are

consistent with NRC GL 86-10 and GL 88-12. Removal of Fire

Protection Program (FPP) requirements does not affect any fire

protection equipment nor plant equipment important to safety, or

involve any physical modifications to plant structures, systems or

components, and therefore is not associated with an accident

initiator or accident mitigator and can not affect the probability

of occurrence of an accident or increase the consequences of an

accident. The licensee controlled Technical Requirements Manual

(TRM) containing the relocated requirements will be maintained in

accordance with TS Section 6.0. ``Administrative Controls'' and

subject to review in accordance with 10 CFR 50.59. Since future

changes to the FPP (i.e., Updated Final Safety Analysis Report and

the TRM) will be evaluated per 10 CFR 50.59, no increase

(significant or insignificant) in the probability or consequences of

an accident previously evaluated will be allowed. Therefore, these

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The changes will not alter the plant configuration (no new or

different type of equipment will be installed) or create changes in

methods governing normal plant operation that will introduce new

failure modes. These changes will not impose different requirements

and proper control of information will be maintained. These changes

will not alter assumptions made in the safety analysis and licensing

basis. Therefore, these changes do not create the possibility of a

new or different kind of accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed changes are administrative in nature and are

consistent with NRC GL 86-10 and GL 88-12. The changes will not

reduce the margin of safety since they have no impact on any safety

analysis assumptions or sequence of events used in any accident

analysis. In addition, any future changes to the FPP will be

evaluated per the requirements of 10 CFR 50.59. Therefore, the

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 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: January 27, 1995.

Description of amendment request: The proposed change to Limerick

Generating Station (LGS) Units 1 and 2 Technical Specifications (TS)

will eliminate the TS active safety function designation of eight

(i.e., four per unit) Drywell Chilled Water System (DCWS) valves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes will eliminate the TS active safety

function designation of eight (i.e., four per unit) DCWS valves. The

DCWS motor operated valves (MOVs) are designated in TS as Primary

Containment Isolation Valves (PCIVs), during operational conditions

(OPCONS) 1, 2, and 3, which mitigate the consequences of design

basis accidents. The proposed changes will prohibit the subject DCWS

valves from opening during OPCONs 1, 2, and 3, thereby,

[[Page 20525]] eliminating the active safety function, and

maintaining a passive safety function. The postulated accidents

which require the Primary Containment to act as a barrier in order

to mitigate the release of radioactivity described in the LGS

Updated Final Safety Analysis Review [Report] (UFSAR) Section 15,

are not affected by these changes. Therefore, the previously

evaluated postulated on-site and off-site radiological effects of

these accidents will not change.

The DCWS valves will be prohibited from opening during OPCONs 1,

2, and 3 by physical changes made to the electrical control

circuitry and administrative controls. Therefore, the probability of

the valves to fail in the open position will diminish, and the

required Primary Containment isolation safety function will be

maintained.

Therefore, these proposed changes will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes remove the affected automatic isolation

relays from the DCWS MOVs' circuitry. These changes eliminate any

postulated relay failure effects on the associated control circuits

and electrical power supplies. The proposed changes do not introduce

any new accident initiators or any new valve failure modes not

previously evaluated.

Therefore, these changes will not create the possibility of a

new or different kind of accident from any accidents previously

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed changes will prohibit the opening of the DCWS

valves which provide backup cooling from RECW [reactor enclosure

cooling water] during OPCONs 1, 2, and 3. The RECW System is not the

normal DCWS cooling alignment, is not required as a backup safety

related drywell cooling system, and this backup alignment is not an

automatic function. The proposed changes do not affect the function

or operation of DCWS, and since the proposed changes and

administrative controls ensure the valves will remain closed during

OPCONs 1, 2, and 3, the capability for Primary Containment isolation

is not affected. Therefore, the changes 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: February 22, 1995.

Description of amendment request: The proposed change to Limerick

Generating Station (LGS) Units 1 and 2 Technical Specifications (TS)

revises various TS Surveillance Requirements to clarify the Emergency

Diesel Generator acceptable steady state voltage range.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed Emergency Diesel Generator steady state voltage

range of 4280 [plus or minus] 120 volts provides voltages through

the 4160V and 480V distribution systems which are within the

operating range of the connected equipment and power system

components. Therefore, the reduced steady state voltage range will

not cause the malfunction of any equipment or affect the operation

of any equipment in a manner which would increase the probability of

occurrence of an accident previously evaluated in the [Safety

Analysis Report] SAR.

Reducing the Emergency Diesel Generator steady state voltage

range in the Technical Specifications maintains the capability of

the Emergency Diesel Generator to start and attain rated voltage and

frequency within 10 seconds and to accept the engineered safeguard

loads in the required time in order to mitigate the consequences of

an accident. The Emergency Diesel Generator automatic voltage

regulator setting is calibrated to within a range of 4266.5 volts to

4308.5 volts. A review of results from recent monthly Emergency

Diesel Generator Surveillance Tests has confirmed that the voltage

regulators currently maintain the Emergency Diesel Generator steady

state voltage within the 4280 [plus or minus] 120 volt range to be

included in the Technical Specifications. Establishing, via

Technical Specification surveillance requirements and administrative

limits within Station Surveillance Test Procedures, that the

Emergency Diesel Generator voltage regulator is maintaining the

steady state voltage within a narrower range (within the existing

range) provides increased assurance that connected equipment

required to mitigate the consequences of an accident will operate as

required.

Therefore, the proposed TS changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Reducing the Emergency Diesel Generator steady state voltage

range in the Technical Specifications to a range of 4280 [plus or

minus] 120 volts does not create any new accident initiators or

affect any existing accident initiators such that a different type

of accident than previously evaluated could result. The function and

operation of the Emergency Diesel Generators and their connected

loads are not changed in a manner which would create the possibility

of an accident of a different type than any previously evaluated.

Therefore, the proposed TS changes do not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

Reducing the Emergency Diesel Generator steady state voltage

range in the Technical Specifications to a range of 4280 [plus or

minus] 120 volts does not reduce the margin of safety. The reduced

range provides increased assurance that the equipment powered by the

Emergency Diesel Generators will start and operate as designed in

order to perform their design basis functions.

Therefore, the proposed TS 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 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: March 1, 1995.

Description of amendment request: The proposed changes will clarify

the concentrations of calibration gas required to calibrate the

Hydrogen and Oxygen Analyzers, and support the requirements of Limerick

Generating Station (LGS) Transient Response Implementation Plant (TRIP)

T-102, ``Primary Containment Control Bases.''

Basis for proposed no significant hazards consideration

determination: [[Page 20526]] 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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed TS [Technical Specification] changes remove

calibration of the H2/O2 Analyzers using zero volume percent

hydrogen (H2) and 100% bottled Nitrogen (N2). A calibration gas

containing zero volume percent H2 and 100% bottled N2 is not

required for calibration of the analyzers to the required accuracy.

Calibration of the H2/O2 Analyzers is done in accordance with the

manufacturer's instructions. The proposed TS changes also revise the

span gas concentration from 5% to 7% to support the requirements of

TRIP T-120. The H2/O2 Analyzers provide indication of the

concentrations of combustible gases in the primary containment and

provide annunciation when combustible gas concentrations reach

unacceptable levels. Failure of the analyzers is not an accident

initiator. The analyzers do not connect to the reactor coolant

pressure boundary; therefore, they do not increase the probability

of a LOCA [loss-of-coolant accident]. The proposed TS changes do not

involve any design changes to analyzers. Therefore, these TS changes

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The H2/O2 Analyzers provide indication and alarms for H2 and O2

concentrations in containment. No physical or design changes to the

analyzers are being made by these TS changes. During normal

operations, the potential for an explosive atmosphere is negligible

due to the absence of H2 sources. For Post-LOCA, conditions the

levels of H2 and O2 in containment have already been evaluated in

LGS UFSAR [Updated Final Safety Analysis Report] Section 6.2.5. No

physical or design changes which could introduce a new analyzer

failure mode are being made. The failure modes of the analyzers are

evaluated in UFSAR Table 6.2-21. Therefore, these TS changes will

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

from any accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

These TS changes will clarify statements in the LGS UFSAR and TS

concerning calibrated ranges of the analyzers. The change of the

span gas from 5% to 7% falls within conditions previously analyzed.

The Bases for TS 3/4.3.7.5 and 3/4.6.6 require operable H2/O2

Analyzers to ensure the analyzers will be available for monitoring,

assessing and controlling H2 and O2 in containment following a LOCA.

These TS changes do not adversely affect operability of the

analyzers or their availability for use during Post-LOCA conditions;

therefore, the margin of safety is not reduced.

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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz.

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

Generating Station, Salem County, New Jersey

Date of amendment request: December 15, 1994.

Description of amendment request: In accordance with 10CFR50.90,

PSE&G proposes to remove Technical Specification Surveillance

Requirement 4.8.1.1.2.h.1, and utilize plant- controlled programs to

govern diesel generator maintenance. To ensure procedural consistency

and reduce the impact of this change on Hope Creek procedures, the

remaining Surveillance Requirements of Technical Specification

4.8.1.1.2.h are not renumbered.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

The proposed change is consistent with the improved Standard

Technical Specifications (NUREG-1433) and does not result in any

changes to the existing plant design. The Hope Creek preventative

maintenance program will utilize diesel generator performance

history, engineering analyses and manufacturer's recommendations as

appropriate for determining diesel generator inspection

requirements. Since the changes do not impact the ability of the

diesel generators and the AC electrical power sources to perform

their function, the changes do not result in a significant increase

in the consequences of any accident previously evaluated. The diesel

generators will continue to function as designed. Therefore, the

proposed change will not impact the probability of occurrence of any

accident previously evaluated.

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

accident from any accident previously evaluated.

This request does not result in any change to the plant design

nor does it involve a significant change in current plant operation.

The diesel generators will be inspected utilizing diesel generator

operating history, engineering analyses and manufacturer's

recommendations as appropriate, and the remaining surveillance

requirements will not be changed. As a result, no new failure modes

will be introduced, and the proposed changes will not create the

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

previously evaluated.

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

safety. The proposed request does not adversely impact the

reliability of the diesel generators. As stated above, the diesel

generator operating history, engineering analyses and the

manufacturer's recommendations will be utilized as appropriate to

perform the diesel generator inspections. In addition, the diesel

generators will continue to perform their design functions. This

request does not involve an adverse impact on diesel generator

operation or reliability. Since the diesel generator function is not

affected by the proposed changes, this request does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070.

Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and

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

NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,

Unit 1, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 95-09).

Description of amendment request: The proposed change would revise

Operating Condition 2.C.(25) to extend the ice condenser Surveillance

4.6.5.1.d to October 1, 1995, to coincide with the Unit 1 Cycle 7

refueling outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not: [[Page 20527]]

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change is temporary and allows a one-time extension

of the ice condenser Surveillance Requirement 4.6.5.1.d for Cycle 7

to allow surveillance testing to coincide with the seventh refueling

outage. The proposed surveillance interval extension will not cause

a significant reduction in system reliability nor affect the ability

of the system to perform the design function. Current monitoring of

plant conditions and continuation of the surveillance testing

required during normal plant operation will continue to be performed

to ensure conformance with TS operability requirements. Therefore,

this change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

from any previously analyzed.

Extending the surveillance interval for the performance of ice

condenser testing will not create the possibility of a new or

different kind of accident. No change is required to any system

configuration, plant equipment, or analyses.

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

The safety limits assumed in the accident analyses and the

design function of the equipment required to mitigate the

consequences of postulated accidents will not be changed

significantly. Existing analysis indicates that the potential

reduction in ice weight resulting from the proposed extension will

continue to maintain the maximum containment accident pressure below

12 pounds per square inch gauge. The ice condenser will continue to

support accident mitigation functions although some Row 1 baskets

could drop slightly below the required 993-pound analysis limit.

Therefore, the plant will be maintained with acceptable ice weights

for accident mitigation and the proposed extension will not

significantly reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 95-11).

Description of amendment request: The proposed change would

relocate the constant numerical value found in the overtemperature

delta temperature and overpower delta temperature equations of

Technical Specifications Table 2.2-1 and place them in the Core

Operating Limits Report (COLR). This would be accomplished by revising

notes 1 and 2 in Table 2.2-1 to state that the values are located in

the COLR. The values of the constants, however, would not be changed.

Also, the ``Overtemperature and Overpower Delta Temperature Setpoint

Parameter Values for Specification 2.2.1'' would be added to the list

of core operating limits specified in Section 6.9.1.14 that are

required to be included in the COLR. In addition, a reference to WCAP-

8745-P-A, ``Design Bases for the Thermal Overpower delta-T and Thermal

Overtemperature delta-T Trip Functions,'' would be added to Section

6.9.1.14.a.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes will allow changes to the constant

numerical values for the overtemperature delta temperature

P(OT[delta-]T) and overpower delta temperature (OP[delta-]T)

equations in accordance with the 10 CFR 50.59 requirements. This

revision does not revise these constants, but relocates them to the

core operating limits report (COLR) that is governed by the 10 CFR

50.59 requirements. The addition of the lag compensator functions

for measured [delta-]T and average temperature in these equations

does not alter the setpoint because this lag function has a value of

unity. Therefore, the proposed revision does not alter plant

functions or setpoints, but does allow for a more timely revision

process for parameters that may require changes due to specific fuel

cycle requirements or updated plant analyses. The use of the lag

functions and revisions to the constant numerical values will be

maintained within the safety analysis for the plant by the 10 CFR

50.59 process requirements. The probability of an accident is not

increased because the plant functions are not altered by the

proposed revision and future changes will be in accordance with 10

CFR 50.59. Additionally, the consequences of an accident are not

increased because the mitigation functions of the OT[delta-]T and

OP[delta-]T functions are not changed and revisions to the equations

that derive the setpoints will be processed under the requirements

of the 10 CFR 50.59 program.

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

from any previously analyzed.

The proposed revision will not change plant functions and future

revisions will continue to be controlled in accordance with the 10

CFR 50.59 requirements. The addition of the lag functions does not

create a new accident potential because these functions have already

been considered in the analysis as shown in NUREG 1431. Therefore,

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

created by the proposed revision.

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

Plant parameters are not altered by the proposed revision and

the OT[delta-]T and OP[delta-]T functions will not reflect a change

in setpoint generation or value. The proposed change will allow

revision of the constant numerical values and use of the lag

compensator functions in accordance with the 10 CFR 50.59 provisions

to ensure the design basis of the plant is maintained. This revision

does not result in changes that reduce the margin of safety because

the OT[delta-]T and OP[delta-]T functions remain unchanged and

future revisions to these functions will be performed in accordance

with 10 CFR 50.59.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 94-15).

Description of amendment request: The proposed change would modify

the Technical Specifications associated with the Post Accident Sampling

(PAS) system by deleting License Condition 2.C.(23)F for Unit 1 and

2.C.(16)g for Unit 2 that require operation of the PAS system in

accordance with referenced letters no later than startup from the

second refueling outage. The submittal also includes a revised

description of operation of the PAS system for insertion into the

Updated Final Safety [[Page 20528]] Analysis Report for staff approval.

This information supersedes the information contained in the letters

referenced in the License Conditions listed above and would be

maintained in accordance with the 10 CFR 50.59 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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change involves the deletion of license conditions

that authorized TVA to operate SQN's postaccident sampling (PAS)

system. TVA proposed change establishes programmatic control of

SQN's PAS system under SQN TS 6.8.4.e and the SQN Final Safety

Analysis Report. Any future changes to SQN's PAS Program would be

governed by the 10 CFR 50.59 process. PAS and analysis will continue

at SQN through grab sample acquisition and laboratory analysis and

will continue to meet the PAS objectives in NUREG-0737, Item II.B.3

and Regulatory Guide 1.97, Revisions 2. Accordingly, the proposed

change does not affect the probability or consequences of an

accident.

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

from any previously analyzed.

The proposed change involves improvements in the operational

reliability of SQN's PAS system by using more reliable laboratory

analysis methods, reducing sampling personnel radiation dose, and

incorporating practical methods for sample acquisition and analysis.

Because the proposed change involves license conditions and sampling

methods that are utilized for postaccident recovery, the potential

for an unanalyzed accident is not created. Consequently, no new

failure modes are introduced.

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

Plant safety margins are established through limiting conditions

of operation, limiting safety system settings, and safety limits

specified in the TSs. As a result of the proposed amendment, there

will be no change to either the physical design of the plant or to

any of these settings and limits. The proposed changes do not affect

the safe operation of SQN. Therefore, there are no changes to any of

the margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 94-18).

Description of amendment request: The proposed change would revise

Surveillance Requirement (SR) 4.0.5 by replacing the current Inservice

Inspection program requirements with the requirements stated in the

Standard Technical Specifications (NUREG-1431). As a result, SR 4.0.5

would more clearly specify the inservice inspection (ISI) program

requirements and the inservice testing (IST) program requirements of

the American Society of Mechanical Engineers (ASME) Code Class 1, 2,

and 3 components. The licensee also proposed deletion of Technical

Specification 3/4.4.10, ``Structural Integrity ASME Code Class 1, 2 and

3 Components,'' and its related Bases information.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Operation of the facility in accordance with the proposed

amendment would not involve any increase in the probability of

occurrence or consequences of an accident previously evaluated. The

Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a

are described in the TSs. The proposed amendment, in accordance with

NUREG-1431 and draft NUREG-1482 permits relief from an American

Society of Mechanical Engineers (ASME) code requirement in the

interim between the time of submittal of a relief request and NRC

approval of the relief. The changes being proposed do not affect

assumptions contained in plant safety analyses or change the

physical design and/or operation of the plant, nor do they affect

TSs that preserve safety analysis assumptions. Any relief from the

approved ASME Section XI code requirements that is implemented prior

to NRC review and approval will require evaluation under the 10 CFR

50.59 process to determine that no TS changes or unreviewed safety

questions exist. This evaluation process will ensure that the impact

of any code relief is thoroughly evaluated and that the structures,

systems, and components remain in conformance with assumptions made

in the safety analysis. The proposed change to delete SQN TS 3/

4.4.10, Structural Integrity, does not affect plant safety analyses

or change the physical design or operation of the plant. The

proposed amendment relocates the structural integrity requirements

under the existing TS Surveillance Requirement (SR) 4.0.5 to allow

these requirements to be governed and controlled within the

inservice inspection (ISI) program. Therefore, operation of the

facility in accordance with the proposed amendment would not affect

the probability or consequences of an accident previously evaluated.

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

from any previously analyzed.

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

Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a

are described in the TSs. The proposed amendment, in accordance with

NUREG-1433 and draft NUREG-1482, permits relief from an ASME code

requirement in the interim between the time of submittal of a relief

request and NRC approval of the relief. The changes being proposed

will not change the physical plant or the modes of operation defined

in the Facility License. The changes do not involve the addition or

modification of equipment nor do they alter the design or operation

of plant systems. Any relief from the approved ASME Section XI code

requirements that is implemented prior to NRC review and approval

will require evaluation under the 10 CFR 50.59 process to determine

that no TS changes or unreviewed safety questions exist. This

evaluation process will ensure that the impact of the code relief is

thoroughly evaluated and that the structures, systems, and

components remain in conformance with assumptions made in the safety

analysis. The proposed change to delete SQN TS 3/4.4.10 does not

affect plant safety analyses or change the physical design or

operation of the plant. The proposed amendment relocates the

structural integrity requirements under the existing TS SR 4.0.5 to

allow these requirements to be governed and controlled within the

ISI program. Therefore, operation of the facility in accordance with

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

different kind of accident from any accident previously evaluated.

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

Operation of the facility in accordance with the proposed

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

[[Page 20529]] The Inservice Inspection and Testing Programs,

pursuant to 10 CFR 50.55a, are required by the SQN TSs. The proposed

amendment, in accordance with NUREG-1431 and draft NUREG-1482

permits relief from an ASME code requirement in the interim between

the time of submittal of a relief request and NRC approval of the

relief. Any relief from the ASME Section XI code is required to be

evaluated under the 10 CFR 50.59 process to determine that no TS

changes or unreviewed safety questions exist. This evaluation

process will ensure that code relief does not affect the ability of

structures, systems, or components to perform their design function,

affect compliance with any TS requirements or reduce the margin of

safety. The proposed change to delete SQN TS 3/4.4.10 does not

affect plant safety analyses or change the physical design or

operation of the plant. The proposed amendment relocates the

structural integrity requirements under the existing TS SR 4.0.5 to

allow these requirements to be governed and controlled within SQN's

ISI program. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a reduction in the margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 95-06).

Description of amendment request: The proposed change would delete

the technical specification requirement that limits and controls loads

traveling over the spent fuel pool (Specification 3.9.7), the graph

that relates the Load Carried Over the Shield to the Allowable Height

Above the Shield Surface (Figure 3.9-1), the crane interlocks and

physical stops surveillance requirements (Specifications 4.9.7.1 and

4.9.7.2), and the related Bases information. These requirements would

be relocated to administratively controlled procedures.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed TS change involves the relocation of a requirement

that does not pertain to limitations or conditions of reactor

operation or to equipment to mitigate design basis accidents or

transients. SQN is proposing to relocate this TS based on NRC's

final policy statement on TS improvement (58 FR 39132, dated July

22, 1993). Based on this criteria, the spent fuel pit (SFP) crane

travel is not important to operational safety and may be relocated

to administratively controlled procedures. By placing the crane

travel requirements in administratively controlled procedures,

adequate controls will remain in place to prevent heavy loads from

traveling over fuel assemblies in the SFP. The administratively

controlled procedure that controls the by-passing of the interlocks

and physical stops is subject to the requirements of TS 6.5.1A.

Therefore, the relocation of this TS will not involve an 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 previously analyzed.

The proposed change involves relocating TS requirements to

another administratively controlled document. No modifications to

the plant are involved. Additionally, there are no changes to the

operation of the plant or equipment proposed. Based on this, the

relocation of this TS will not create the possibility of a new or

different kind of accident from any previously analyzed.

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

The proposed change involves the relocation of TS requirements

to administratively controlled procedures. The relocation of this

requirement is based on the criteria endorsed in the Commission's

Policy Statement on TS improvements as it pertains to 10 CFR 50.36.

Additionally, this change does not alter the basic design and safety

analysis requirements, as discussed in the Updated Final Safety

Analysis Report. Therefore, the deletion of this TS will not involve

a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 94-19).

Description of amendment request: The proposed change would revise

the Action statement for Technical Specification 3.8.1.1 by inserting a

new Action a, relabeling and modifying existing Action a to become

Action b, adding a footnote referenced to Action b, renumbering the

subsequent action statements, and adding information to the Bases that

amplifies the action statements. The proposed new Action a would no

longer address required actions for diesel generator testing. It would

require that, should one of the AC electrical power sources listed be

inoperable, then operability of the remaining offsite AC circuit be

demonstrated by performing Surveillance Requirement 4.8.1.1.1.a within

1 hour and at least once per 8 hours thereafter. If two offsite

circuits cannot be restored within 72 hours, the unit must be placed in

hot standby within the next 6 hours and in cold shutdown within the

following 30 hours.

The proposed change to Action b would address the testing

requirements should a diesel generator become inoperable. It would

require testing of operable diesel generators if the inoperability of

the affected diesel generator has the potential to be the result of a

common cause failure. A footnote would clarify that the common cause

determination must be completed regardless of how long the diesel

generator inoperability persists or Surveillance 4.8.1.1.2.a.4 must be

completed to verify diesel generator operability. The proposed change

to the Bases would provide clear guidance on the use of common cause

failure determinations to eliminate unnecessary diesel generator

testing and would define when testing is required to verify diesel

generator operability.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined [[Page 20530]] that it does not represent

a significant hazards consideration based on criteria established in

10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in

accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed revisions do not alter the plant features or

operating practices. This revision will reduce unnecessary starts of

the diesel generators (D/Gs) when a common cause failure is not

involved or for an inoperable offsite circuit. This change will not

affect the accident mitigation capabilities of the D/Gs, but should

improve the reliability by reducing the wear and tear associated

with starting the D/Gs. The D/Gs are not the source of a postulated

accident and because this change does not alter plant functions or

operating practices the probability of an accident is not increased.

The D/G's operability will continue to be verified for conditions

that indicate a potential common-cause failure to ensure accident

mitigation capabilities are not affected. Therefore, this revision

will continue to provide actions that will support alternating-

current (ac) electrical power source safety functions without

unnecessary degradation of the D/Gs and will not increase the

consequences of an accident.

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

from any previously analyzed.

The D/Gs are not the source of accidents and the proposed

revision will not alter plant functions or actions by more

appropriately limiting the conditions when a D/G must be verified

operable. Therefore, the possibility of a new or different accident

is not created

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

This revision does not alter plant functions that provide the

margin of safety. The reduction of D/G testing will only be allowed

for situations where the operable D/Gs are not affected by the

conditions resulting in the ac power source inoperability. This

reduced testing should improve D/G reliability for accident

mitigation functions and further ensure the margin of safety

provided by the D/Gs. Therefore, the margin of safety is not reduced

by the proposed revision to limit unnecessary D/G starts.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 95-02).

Description of amendment request: The proposed change would add

Limiting Condition for Operation 3.0.6 to allow equipment that has been

removed from service or declared inoperable to be returned to service

under administrative control in order to perform testing required to

demonstrate operability. It would be applicable for operability testing

of the inoperable equipment or other equipment that requires the

operability feature to be in service in order to perform the test. A

related change to the Bases would provide amplifying explanation on the

use of this new provision. In addition, a proposed change to Action 18

of Table 3.3-3, ``Engineered Safety Feature Actuation System

Instrumentation,'' would clarify the time interval that an instrument

channel may be in the bypass condition. For those instruments that

reference Action 18, the change would allow the bypass for 6 hours.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The addition of the exception to TSs 3.0.1 and 3.0.2 and the

definition for the time to place a channel in bypass will not change

plant equipment or the operating practices at SQN. The exception

will allow testing to be performed with inoperable equipment

returned to service under administrative controls, but will not

change functions. The function will be available from other

redundant channels during the brief durations that the new provision

would be utilized. The specified time interval to achieve a bypass

condition will clarify the implementation of the action requirement

with the affected functions remaining available through the

redundant operable channels. This clarification does not change the

intent of the action but does set the previously undefined time

interval.

The proposed change affects actions associated with the

actuation of functions to mitigate accidents and are not the source

of an accident. Therefore, the probability of an accident is not

increased. The affected functions provide accident mitigation

functions and the proposed revisions serve to ensure equipment can

be maintained in required conditions within acceptable time

intervals and administrative controls. The brief periods utilized

for the TSs 3.0.1 and 3.0.2 exception will not significantly affect

the accident mitigation capabilities because of the availability of

redundant equipment. In addition, the benefit of performing

operability testing to return equipment permanently to service or to

maintain the operability of other equipment outweighs the slight

reduction in safety function actuation redundancy. Therefore, the

proposed change will not significantly increase the consequences of

an accident.

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

from any previously analyzed.

The proposed changes affect functions utilized to mitigate an

accident and are not the source of an accident. The exception

provides reasonable flexibility to maintain equipment operability

and the bypass time interval reduces the potential for damage to

safety related equipment. Because plant functions are not changed as

a result of this request the possibility of a new or different kind

of accident is not created.

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

The proposed change does not alter setpoints or operating

considerations that maintain the margin of safety for SQN. These

changes provide needed flexibility to perform TS required testing

and clarifications for implementing action requirements. These

changes will slightly affect the redundancy of the affected safety

functions but provide greater benefit for maintaining equipment in

an operable condition. Therefore, the margin of safety provided by

the affected equipment has not changed and the proposed change will

not result in a reduction.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 94-04).

Description of amendment request: The proposed amendment would

change the power range neutron flux channel calibration frequency

[[Page 20531]] surveillance requirement from monthly to every 31

effective full power days and delay the requirement to perform the

surveillance for 96 hours after reaching 15 percent power. A proposed

change to the Bases would provide amplifying information.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The likelihood that an accident will occur is neither increased

or decreased by this TS change, which only affects when the first

surveillance is performed following an outage and changes the

frequency of performance of the surveillance. Before start-up

following refueling outage, the power range high trip setpoint is

set below 85 percent power, typically 60 percent, for conservatism.

The power range low trip setpoint is set at 22 percent power, TS

requires the setpoint to be less than or equal to 25 percent power.

These settings are in addition to the conservatism built into start-

up following a refueling outage. Therefore, delaying the first

performance for 96 hours will not impact on the operation of the

plant since the setpoints are set conservatively. Also, the change

of the frequency to every 31 effective full power days (EFPD) only

delays the surveillance when the plant is operated at reduced power.

During operation at reduced power changes in the neutron flux are

also reduced. Therefore, changing the frequency from monthly to

every 31 EFPD allows slow changes in neutron flux during the fuel

cycle to be more accurately detected and evaluated.

This TS change will not impact the function or method of

operation of plant equipment. Thus, there is not a significant

increase in the probability of a previously analyzed accident due to

this change. No systems, equipment, or components are affected by

the proposed change. Thus, the consequences of a malfunction of

equipment important to safety previously evaluated in the Updated

Final Safety Analysis Report are not increased by this change.

The proposed changes provide TS improvements that ensure the

system operates within the bounds of SQN's accident analysis as

contained in the Final Safety Analysis Report (FSAR) and only

affects when a surveillance is performed. This change has no impact

on accident initiators and does not involve a physical modification

to the plant. Accordingly, the proposed changes do not involve an

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 previously analyzed.

This revision will not change any plant equipment, system

configurations, or accident assumptions. This change will more

accurately monitor changes in the condition of the core.

Fuel burn-up is necessary to change the relationship between the

incore axial power and the excore detectors response. At reduced

levels the effectiveness of the monitoring activity is reduced.

Therefore, changing the frequency to 31 EFPD allows slow changes in

neutron flux during the fuel cycle to be more accurately detected

and evaluated. Delaying the first performance of the surveillance

requirement, until 96 hours after reaching 15 percent rated thermal

power, will allow the unit to be in a more stable condition.

Therefore, this change will not affect the safety function of any

components and will create the possibility of a new or different

kind of accident.

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

The proposed changes provide TS improvements for SQN's power

range monitoring system that ensure the system operates within the

bounds of SQN's accident analysis as contained in the FSAR since

only the time interval between performances of the surveillance is

being extended. This change does not involve a physical modification

to SQN's power range monitoring system. Accordingly, the margin of

safety has not been reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 95-08).

Description of amendment request: The proposed change would (1)

change the core alteration definition to limit the term to reactor

vessel internal activities that could have an affect on core

reactivity, (2) change the quadrant power tilt ratio definition to

eliminate the conflict in the definition of the term and its use in

Surveillance Requirement 4.2.4.2, and (3) revise the Unit 1 Operational

Modes parameters in Table 1.1 to be consistent with the description in

Table 1.1 for Unit 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The proposed

changes provide TS improvements that ensure the plant operates

within the bounds of SQN's accident analysis as contained in the

Final Safety Analysis Report (FSAR) and only affects the definitions

and does not have any affect on any work performed. The change to

core alteration is to clarify those components that may result in

reactivity changes. The change will not effect movement of fuel or

components that effect reactivity, therefore, a fuel handling

accident will not be effected. The change in the definition of

quadrant power tilt ratio (QPTR) allows the alternate method of

determining QPTR to be utilized. The current TS surveillance

requirement (SR) and bases allow alternate means for determining

QPTR, therefore, revising the definition will have no effect on any

accident. The revision to the mode parameters is administrative in

nature, therefore it will have no effect on any accident. This

change has no impact on accident initiators and does not involve a

physical modification to the plant. Accordingly, the proposed

changes do not involve an 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 previously analyzed.

This revision will not change any plant equipment, system

configurations, or accident assumptions. This change will better

define the associated parameters and will eliminate potential

ambiguity and confusion. The change in the definition of core

alteration allows components that do not affect reactivity to be

moved within the reactor vessel. The change in the definition will

not effect the monitoring of QPTR with one channel inoperable. The

core will be monitored in accordance with the SRs. Therefore, this

change will not affect the safety function of any components and

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

accident.

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

The proposed changes provide improvements for SQN's TS. This

change does not involve a physical modification to the plant nor

change the methods of monitoring plant parameters. Accordingly, the

margin of safety has not been reduced.

[[Page 20532]] The NRC has reviewed the licensee's analysis and,

based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: March 24, 1995.

Description of amendment request: This amendment request proposes

to revise Technical Specification (TS) 1.7, ``Containment Integrity,''

TS 3/4.6.1, ``Containment Integrity,'' TS 3/4.6.3, ``Containment

Isolation Valves,'' and the associated Bases. These proposed changes

will relocate TS Table 3.6-1, ``Containment Isolation Valves,'' to Wolf

Creek Generating Station procedures. This proposed change is in

accordance with the guidance provided in Generic Letter 91-08,

``Removal of Component Lists from Technical Specifications,'' dated May

6, 1991.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes simplify the technical specifications, meet

the regulatory requirements for control of containment isolation,

and are consistent with the guidelines of GL 91-08. The procedural

details of Technical Specification Table 3.6-1 have not been

changed, but only relocated to a different controlling document. The

proposed changes are administrative in nature, should result in

improved administrative practices, and do not affect plant

operations.

The probability of occurrence of a previously evaluated accident

is not increased because this change does not introduce any new

potential accident initiating conditions. The consequences of an

accident previously evaluated is not increased because the ability

of [the] containment to restrict the release of any fission product

radioactivity to the environment will not be degraded by this

change.

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

or different kind of accident from any 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, this proposed change will not

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

any previously evaluated.

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

in the margin of safety.

The administrative change to relocate Technical Specification

Table 3.6-1 to appropriate plant procedures does not alter the basic

regulatory requirements for containment isolation and will not

adversely affect containment isolation capability for credible

accident scenarios. Adequate control of the content of the table is

assured by existing plant procedures.

The proposed relocation of Technical Specification Table 3.6-1

does not alter the requirements for containment isolation valve

operability currently in the technical specifications. The LCO

[limiting condition for operation] and Surveillance Requirements

would be retained in the revised technical specifications.

Therefore, the proposed change will not affect the meaning,

application, and function of the current technical specification

requirements for the valves in Table 3.6-1.

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 locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 4, 1995, as supplemented by letter

dated April 5, 1995.

Description of amendment request: The proposed amendment would

change the technical specifications on moderator temperature

coefficient. The proposed change constitutes a one time deviation not

to perform the two-thirds end-of-cycle moderator temperature

coefficient test for Cycle 7.

Date of individual notice in the Federal Register: April 11, 1995

(60 FR 18432).

Expiration date of individual notice: May 11, 1995.

Local Public Document Room location: University of New Orleans

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

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

[[Page 20533]] prepared an environmental assessment under the special

circumstances provision in 10 CFR 51.12(b) and has made a determination

based on that assessment, it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application for amendment: February 24, 1995.

Brief description of amendment: The amendment revises Technical

Specification Section 4.6.1.2.a, Primary Containment/Containment

Leakage, to reference 10 CFR Part 50, Appendix J, as modified by

approved exemptions.

Date of issuance: April 10, 1995.

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

30 days.

Amendment No.: 183.

Facility Operating License No. DPR-69: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 8, 1995 (60 FR

12789).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 10, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: September 30, 1994, as

supplemented on March 24, 1995.

Brief Description of amendments: The proposed change will revise

Technical Specification requirements to eliminate the reactor scram and

isolation functions of the Main Steam Line Radiation Monitors. The

March 24, 1995, supplement provided clarifying information only and did

not affect the NRC's determination of no significant hazards

considerations.

Date of issuance: March 31, 1995.

Effective date: March 31, 1995.

Amendment No.: 160.

Facility Operating License Nos. DPR-71 and DPR-62. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55867). The March 24, 1995, submittal provided clarifying information

only and did not affect the no significant hazards consideration as

published in the Federal Register.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 31, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

Date of application for amendment: July 22, 1994, as supplemented

March 6, 1995.

Brief description of amendment: The amendments change the Technical

Specifications to implement a performance based assessment program,

including corresponding organizational and functional changes.

Specifically, the changes affect the independent assessment of plant

activity and the independent review function, the independent

assessment of plant activity and the Independent Safety Engineer

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