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

Federal RegisterMay 25, 1994

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 2, 1994, through May 13, 1994. The last

biweekly notice was published on May 12, 1994 (59 FR 24745).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

delivered to Room 6D22, Two White Flint North, 11555 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

20555. The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By June 24, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

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

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

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

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

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the

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

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

the Commission by a toll-free telephone call to Western Union at 1-

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

operator should be given Datagram Identification Number N1023 and the

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

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

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC 20555, and at the local public document

room for the particular facility involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: March 25, 1994.

Description of amendments request: The proposed amendment would

make the following administrative changes to the Technical

Specifications.

Brunswick Unit 1

1. Bases Section 2.2.1: Remove references to the Rod Sequence

Control System (RSCS) in item 2 on page B 2-4.

2. Bases Section 2.2.1: Correct typographical error in acronym

for hydrogen water chemistry in item 6 on page B 2-6.

3. TS 3.1.4.1: Correct typographical errors in action d,

misspelling of preset, and action d.1, misspelling of BPWS acronym,

on page 3/4 1-14.

4. TS Table 4.3.4-1: Remove references to the RSCS in item g of

the Notes on page 3/4 3-52.

5. TS Table 3.3.5.5-1 Label each item to permit identification

consistent with the scheduling system used for surveillance testing

on pages 3/4 3-64a.

6. TS Table 4.3.5.5-1 Label each item to permit identification

consistent with the scheduling system used for surveillance testing

on page 3/4 3-64c.

7. TS 4.3.6.1.1: Correct typographical error that references

Non-existent Table 4.3.6.1.1-1 to provide correct reference of Table

4.3.6.1-1 on page 3/4 3-88.

8. TS 3.4.2: Correct typographical error indicating extraneous

second footnote on page 3/4 4-4.

Brunswick Unit 2

1. TS Table 2.2.1-1: Correct typographical error in item 2.b

under allowable values by changing 115% to 115.5% on page 2-4.

2. Bases Section 2.2.1: Remove references to the Rod Sequence

Control System (RSCS) in item 2 on page B 2-4.

3. Bases Section 2.2.1: Remove references to the Rod Sequence

Control System in item 10 and revise bases description of the Select

of the Select Rod Insertion consistent with removal of the RSCS on

pages.

4. TS 3.1.4.1: Correct typographical error in action d.1 to

correct misspelling of BPWS acronym on page 3/4 1-14.

5. TS Table 4.3.1-1: Correct grammatical omission of the word

``is'' in item e of the Notes on page 3/4 3-9.

6. TS Table 4.3.1-1: Remove references to the RSCS in item g of

the Notes on page 3/4 3-52.

7. TS Table 3.3.5.5-1: Label each item to permit identification

consistent with the scheduling system used for surveillance testing

on page 3/4 3-64a.

8. TS Table 4.3.5.5-1: Label each item to permit identification

consistent with the scheduling system used for surveillance testing

on page 3/4 3-64c.

9. TS 3.3.6.2: Eliminate footnote, revise applicability

statement and correct typographical errors in actions d and e that

references non-existent Specification on page 3/4 3-93.

10. Base Section 3/4.1.4: Correct identification of Reference

cited to reference 6 on page B 3/4 1-4.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated because the proposed change [sic] is

administrative in nature. These changes do not alter the

configuration or operation of the facility. The Limiting Safety

Systems Settings and Safety Limits specified in the current

Technical Specifications remain unchanged.

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

new or different kind of accident from any accident previously

evaluated. The safety analysis of the facility remains complete and

accurate. There are no physical changes to the facility and the

plant conditions for which the design basis accidents have been

evaluated are still valid. The operating procedure and emergency

procedures are unaffected with the possible exception of resolving

special notations that may have recognized the typographical errors

that are being corrected.

3. The margins of safety are established through the Limiting

Conditions of Operation, Limiting Safety Systems Settings and Safety

Limits specified in the Technical Specifications. Since there are no

changes to the physical design or operation of the facility, these

margins will not be changed.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

NRC Project Director: William H. Bateman

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck

Plant, Middlesex County, Connecticut

Date of amendment request: January 28, 1994.

Description of amendment request: The proposed amendment will

remove an exception for the purge and vent valves from surveillance

requirement (SR) 4.6.1.2.d and remove SR 4.6.1.2.f.

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 involve an [significant hazards

consideration] SHC because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change modifies SR 4.6.1.2.d. Currently this SR

indicates the purge supply and exhaust valves have an exception from

the 10CFR50 Appendix J, Type B and C tests. The proposed technical

specification change is consistent with current surveillance

procedures and the [Final Safety Analysis Report] FSAR. The second

proposed change, which removes SR 4.6.1.2.f, reflects current

containment leakage surveillance requirements. The present location

of SR 4.6.1.2.f could imply that containment leakage surveillance

requirements are met by performing SR 4.9.9. However, SR 4.9.9 is

applicable only during core alterations or movement of irradiated

fuel and not during the modes when Technical Specification 3.6.1.2

is applicable. These changes have no effect on actual Appendix J

testing of valves or the current plant accident analysis. Therefore,

the proposed changes cannot increase the probability or consequences

of an accident previously evaluated.

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

from any previously analyzed.

The proposed changes do not introduce any new failure modes. The

plant will continue to operate as designed and there will be no

change to the testing of valves. The proposed changes will not

modify the plant response to the point where it can be considered a

new accident. Therefore, the proposed changes will not create the

possibility of a new or different kind of accident form any

previously evaluated.

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

The proposed changes modify SR 4.6.1.2.d which, as presently

written, indicates that the purge supply and exhaust valves are an

exception to the 10CFR50 Appendix J, Type B and C test and

therefore, no exception is required. This is supported by current

surveillance procedures which include the purge supply and exhaust

valves as part of the Type B and C tests. In addition, the proposed

changes are consistent with the FSAR. FSAR Table 7.3-1 ``Containment

Penetrations,'' lists the purge supply and exhaust valves as

required to receive Type B and C tests. Therefore, these proposed

changes revise SR 4.6.1.2.d to reflect actual surveillance

procedures and offer no revisions or reductions to current

surveillance testing. Therefore, these changes will not result in a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: April 13, 1994

Description of amendment request: The proposed amendment request

would revise the Technical Specifications to amend Sections 3.1.F and

4.13 to allow the repair of steam generator tubes by sleeving as an

alternative to plugging. Additionally, a new tube acceptance criteria,

F*, is proposed which would allow tubes that are degraded in a location

not affecting structural integrity of the tube to remain in service.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with the requirements of 10 CFR 50.92, the

proposed Technical Specification change is deemed to involve no

significant hazards considerations because operation of Indian Point

Unit No. 2 would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated since the integrity

of the steam generator tubes after sleeving will be equivalent to

that of the original tubes. The sleeve, sleeve joint, and F* joint

have been analyzed and tested for design, operating, and faulted

condition loadings in accordance with NRC Regulatory [G]uide 1.121

safety factors. The potential for a tube rupture is not increased

with sleeving or F*. At worst case, a tube leak would occur,

resulting in a small primary to secondary leak. Primary to secondary

leakage occurring from within the sleeved or F* portions of the tube

is bounded by the steam generator tube rupture scenario evaluated in

the Final Safety Analysis Report. In addition, the steam generator

tube remains capable of performing its required heat transfer

function. Placing a sleeve in the steam generator tube or leaving a

tube in service with a defect in a portion of the tube that provides

no function results in a more efficient steam generator than

plugging an affected tube. Thus, the consequences of any accident

previously evaluated are not increased because the structural

integrity and the heat transfer capability of the steam generators

are not significantly altered by the proposed change.

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

accident from any accident previously evaluated because both the

structural integrity and the heat transfer capability of the steam

generators will not be significantly affected by the use of either

of the sleeving processes or the implementation of the F* criteria.

Testing and previous experience indicate that any primary to

secondary leakage would be well below technical specification

limits. In addition, in the unlikely event the defective tube failed

completely at the defect, the remaining sleeve end or F* joint would

restrain tube movement due to the sleeve end geometry or length of

expanded contact within the tubesheet bore. Therefore, there is no

threat to adjacent tubes and no other plant systems will be affected

by this change. Thus, there is no potential for a new or different

kind of accident.

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

heat transfer capabilities of Indian Point 2 Steam Generators will

be improved by utilizing the proposed sleeving process or

implementing the F* criteria rather than the currently required tube

plugging and subsequent loss of heat transfer area. The proposed

change will allow a repaired (sleeved) tube or a tube with a tube

end defect below the F* distance to remain in service, rather than

completely blocking the tube's flow with plugs. Because the

structural integrity of the tubes will be unaltered, the net effect

of implementing the proposed change, rather than the currently

required plugging procedure, will be an increase in the heat

transfer characteristics of the steam generator. Westinghouse has

done an evaluation of selected LOCA [loss-of-coolant accident] and

non-LOCA transients to verify that use of sleeves resulting in a

plugging equivalency at the current plant limit will not have an

adverse affect on the thermal-hydraulic performance of the plant.

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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: November 15, 1991, as supplemented

February 22, March 11, and April 7, 1994.

Description of amendment request: The amendment request, as

submitted November 15, 1991, proposed completely rewritten requirements

for the instrumentation and control (I&C) sections of the Palisades

Technical Specifications (TS) and was initially noticed in the Federal

has updated its submittal, providing (1) changes to pages affected by

intervening amendments, (2) clarifications suggested by NRC and

Palisades reviewers, (3) addition of two instrument channels to the

accident monitoring instruments Limiting Condition for Operation (LCO),

(4) deletion of surveillance requirements for safety injection tank

(SIT) instruments, as suggested by Generic Letter (GL) 93-05, ``Line-

Item Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' and (5) addition of

a general ``Applicability'' LCO which appears in the Standard TS but

not in the Palisades TS. Changes (4) and (5) were not addressed in the

initial proposed no significant hazards consideration (NSH)

determination. The licensee's NSH analysis for these two changes was

provided in its April 7, 1994, letter to the NRC and is discussed

below.

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:

Consumers Power Company finds that activities associated with

the February 22, 1994 and March 11, 1994 Instrument and Control

Technical Specification change revisions include no significant

hazards; and accordingly, a no significant hazards determination in

accordance with 10CFR50.92(c) is justified. The following summary

supports the finding that the proposed change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Neither the deletion of instrument surveillance requirements for

the Safety Injection Tank (SIT) instrumentation nor the addition of

allowance of temporarily returning inoperable equipment to service

for maintenance or testing would affect the probability or

consequences of an accident.

The SIT instrument channels themselves have no accident

function. Their only purpose is to allow verification that the SITs

themselves are operable. Surveillance requirements for these

instruments were purposely deleted from STS during the Technical

Specification Improvement Program. Their removal from Technical

Specifications was suggested in GL 93-05.

Returning inoperable equipment to service as allowed by LCO

3.0.5 is necessary if failed channels are to be restored to operable

status. The restoration of such channels enhances the ability to

monitor for and mitigate abnormal operating conditions and

accidents.

Therefore, the proposed changes do 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 accident previously evaluated.

The proposed changes would not alter the operating conditions of

the plant systems, and would not reduce the reliability of any plant

safety equipment.

Therefore, this change does 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.

The proposed changes would not affect the setpoints, capacities,

or operating limits for any equipment. Therefore, the proposed

changes do not involve a significant reduction of 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, as revised, involves no significant hazards

consideration.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

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

NRC Project Director: Ledyard B. Marsh.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: April 7, 1994.

Description of amendment request: The proposed amendment would

change certain Technical Specifications (TS) to relocate fuel cycle-

specific parameter limits that can generally change with core reloads

to a Core Operating Limits Report (COLR) in accordance with the

guidance of Generic Letter 88-16, ``Removal of Cycle-Specific Parameter

Limits from Technical Specifications.'' Several of the TS bases would

also be revised to refer to limits relocated to the COLR. In each case

where TS limits would be relocated to the COLR, the limits placed in

the COLR would be unchanged and the appropriate bases would be revised

accordingly.

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 following evaluation supports the finding that operation of the

facility in accordance with the proposed TS would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to the TS simply move the values and

parameters for fuel cycle-specific limits from the TS to a Core

Operating Limits Report (COLR). The requirements to maintain the

plant within appropriate bounds are retained in the TS. The values

of the cycle-specific parameter limits in the COLR are determined

using an NRC-approved methodology and remain consistent with all

applicable limits of the plant safety analyses that are addressed in

the Final Safety Analysis Report (FSAR). A requirements for the COLR

and identification of the approved methodology documents are added

to the TS. There are no associated changes in plant operation.

Therefore, operation of the facility in accordance with the proposed

TS would not result in a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any previously evaluated.

As discussed above, the proposed changes do not remove or

alleviate any requirements to maintain the plant within the

appropriate bounds. There are no associated changes in plant

operation. Therefore, operation of the facility in accordance with

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

different kind of accident from any previously evaluated.

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

The proposed changes to the TS simply move the values and

parameters for cycle-specific limits from the Specifications to a

Core Operating Limits Report (COLR). The requirements to maintain

the plant within appropriate bounds are retained in the TS. The

values of the cycle-specific parameter limits in the COLR are

determined using an NRC-approved methodology and remain consistent

with all applicable limits of the plant safety analyses that are

addressed in the Final Safety Analysis Report (FSAR). A requirement

for the COLR and identification of the approved methodology

documents are added to the TS. There are no associated changes in

plant operation. Therefore, operation of the facility in accordance

with the proposed TS would 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

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

NRC Project Director: Ledyard B. Marsh.

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

Michigan

Date of amendment request: March 29, 1994, as corrected April 26,

1994.

Date of amendment request: March 29, 1994, as corrected April 26,

1994.

Description of amendment request: The proposed amendment would

modify the surveillance requirements for scram discharge volume vent

and drain valves and isolation actuation instrumentation and modify the

required actions and surveillance requirements for the emergency diesel

generators to reduce testing during power operation. These changes are

in accordance with guidance contained in Generic Letter (GL) 93-05

``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation,'' dated

September 27, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes to the frequency of testing for these

components will reduce the probability of failure due to wear and

eliminate the possibility of initiating transients during testing of

these components. Therefore, the proposed changes will result in a

decrease in the probability of previously evaluated accidents.

Further, the proposed changes do not alter the design, function, or

operation of the components involved and therefore, do not affect

the consequences of any previously evaluated accident.

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

or different kind of accident from any accident previously

evaluated. As stated above, the proposed changes do not alter the

design, function, or operation of the components involved and

therefore, no new accident scenarios are created.

3. The proposed changes do not involve a significant reduction

in a margin of safety. As developed in Reference 3 [NUREG-1366,

``Improvement to Technical Specification Surveillance

Requirements,'' dated December 1992] and endorsed in Reference 2 [GL

93-05], the proposed changes to the testing frequency will increase

the margin of safety through reduced equipment wear and elimination

of opportunities to induce transients.

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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: Ledyard B. Marsh.

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

Michigan

Date of amendment request: April 26, 1994.

Description of amendment request: The proposed amendment would

relocate tables of instrument response time limits from the Technical

Specifications to the Updated Final Safety Analysis Report (UFSAR) in

accordance with the guidance contained in Generic Letter 93-08 dated

December 29, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes delete and subsequently relocate the details of

Technical Specification Table 3.3.1-2, ``REACTOR PROTECTION SYSTEM

RESPONSE TIMES,'' Table 3.3.2-3, ``ISOLATION ACTUATION SYSTEM

INSTRUMENTATION RESPONSE TIME,'' and Table 3.3.3-3, ``EMERGENCY CORE

COOLING SYSTEM RESPONSE TIMES,'' consistent with the guidance

provided by Generic Letter 93-08 dated, December 29, 1993, entitled,

``Relocation of Technical Specification Tables of Instrument

Response Time Limits.'' Generic Letter 93-08 recommends the removal

and subsequent relocation of various Technical Specification tables

which denote instrument and system response time limits. The

response time limits and associated footnotes are proposed to be

relocated to the Fermi 2 Updated Final Safety Analysis Report

(UFSAR). This allows Fermi 2 to administratively control subsequent

changes to the response time limit tables in accordance with 10 CFR

50.59. The procedures which contain the various response time limits

are also subject to the change control provisions in the

Administrative Controls section of the Technical Specifications. The

proposed change only relocates the existing response time limits.

The Surveillance Requirements and associated Actions are not

affected and remain in the Technical Specifications. Relocating this

information does not affect the initial conditions of a design basis

accident or transient analysis. Since any subsequent changes to the

UFSAR or procedures are evaluated in accordance with 10 CFR 50.59,

no increase in the probability or consequences of an accident

previously evaluated is allowed. Further, the proposed changes do

not alter the design, function, or operation of the components

involved and therefore, do not affect the consequences of any

previously evaluated accident.

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

or different kind of accident from any accident previously

evaluated. The proposed changes will not impose any different

operational or surveillance requirements. The changes propose to

relocate these response time limit tables to other plant documents

whereby adequate control of information is maintained. Further, as

stated above, the proposed changes do not alter the design,

function, or operation of the components involved and therefore, no

new accident scenarios are created.

3. The proposed changes do not involve a significant reduction

in a margin of safety. The proposed change will not reduce a margin

of safety because it has no impact on any safety analysis

assumption. The proposed change does not alter the scope of

equipment currently required to be OPERABLE or subject to

surveillance testing nor does the proposed change affect any

instrument setpoints or equipment safety functions. In addition, the

values to be transposed from the Technical Specifications to the

UFSAR are the same as the exiting Technical Specifications. Since

any future changes to these requirements in the UFSAR or procedures

will be evaluated per the requirements of 10 CFR 50.59, no reduction

in a margin of safety is allowed. Therefore, the change does not

involve a significant reduction in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: Ledyard B. Marsh.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of amendment request: March 30, 1994.

Description of amendment request: The proposed amendments would

allow the analog channel operational test interval for radiation

monitoring instrumentation to be increased from monthly to quarterly.

The proposed amendments are said by the licensee to be consistent with

NRC staff recommendations and guidance contained in NUREG-1366,

``Improvements to Technical Specifications Surveillance Requirements,''

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

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation.''

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 requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Decreasing the frequency of the radiation monitor analog

channel operational test from monthly to quarterly will have no

impact upon the probability of any accident, since the radiation

monitors are not accident initiating equipment. Also, no credit is

taken in accident analyses for automatic actions performed by

radiation monitors contained in Catawba's Technical Specifications,

so the requested amendments will have no adverse impact upon the

consequences of any accident.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated above, the radiation monitors are not accident

initiating equipment. No new failure modes can be created from an

accident standpoint. The plant will not be operated in a different

manner.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. Plant safety margins will be

unaffected by the proposed changes. No safety equipment which is

taken credit for in accident analyses will be affected by the

requested amendments. The availability of the affected radiation

monitors will be increased as a result of the proposed amendments

because the monitors will not have to be made unavailable for

testing as frequently. In addition, radiation monitor operating

experience supports the proposed amendments. Finally, the proposed

amendments are consistent with the NRC position and guidance set

forth in NUREG-1366 and Generic Letter 93-05.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730.

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: David B. Matthews.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: April 19, 1994.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) 4.0.5 a,

``Applicability--Surveillance Requirements.'' The licensee proposes to

delete the wording ``. . . (g), except where specific written relief

has been granted by the Commission pursuant to 10 CFR, Section

50.55a(g)(6)(i)'' in TS 4.0.5 a, for the inservice inspection and

testing programs. With the revisions to the Technical Specifications,

upon finding an ASME Code requirement impractical because of

prohibitive dose rates or limitations in the design, construction, or

system configuration, the licensee may implement the relief request

once it has been submitted to the NRC provided it has been: (1)

Acceptably reviewed pursuant to 10 CFR 50.59; (2) approved by the plant

staff in accordance with the administrative process described in the

inservice inspection and testing programs administrative procedures;

and (3) reviewed and approved by the Plant Nuclear Safety Committee.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments remove the wording ``. . . (g), except

where specific written relief has been granted by the Commission

pursuant to 10 CFR, Section 50.55a(g)(6)(i)'', provided a 10 CFR

50.59 evaluation is performed. The Inservice Inspection and Testing

Programs are described in the Technical Specifications pursuant to

10 CFR 50.55a. In addition, the proposed amendments, in accordance

with NUREG 1431 and draft NUREG 1482, provide relief to the ASME

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

relief request until the NRC has issued a safety evaluation and

granted the relief. The changes being proposed are administrative in

nature and do not affect assumptions contained in plant safety

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

they affect Technical Specifications that preserve safety analysis

assumptions. Any relief from the approved ASME Section XI code

requirements will require a 10 CFR 50.59 evaluation to ensure no

Technical Specification changes or unreviewed safety questions

exist. Therefore, operation of the facility in accordance with the

proposed amendments would not affect the probability or consequences

of an accident previously analyzed.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The changes being proposed are administrative in nature and will

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

the Facility License. The change does not involve the addition or

modification of equipment nor does it alter the design or operation

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

requirements will require a 10 CFR 50.59 evaluation to ensure no

Technical Specification changes or unreviewed safety questions

exist. Therefore, operation of the facility in accordance with the

proposed amendments would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

amendments would not involve a significant reduction in a margin of

safety.

The changes being proposed are administrative in nature and do

not alter the bases for assurance that safety-related activities are

performed correctly or the basis for any Technical Specification

that is related to the establishment of or maintenance of a safety

margin. Any reliefs from the approved ASME Section XI code

requirements will require a 10 CFR 50.59 evaluation to ensure no

Technical Specification changes or unreviewed safety questions

exist. Therefore, operation of the facility in accordance with the

proposed amendments would not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: April 19, 1994.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications by increasing the

surveillance interval specified for air or smoke flow test through the

containment spray header from ``at least once per 5 years'' to ``at

least once per 10 years.'' The licensee stated that the proposed

surveillance interval is consistent with both Generic Letter 93-05,

``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation'' and

NUREG-1366, ``Improvements to Technical Specifications Surveillance

Requirements.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments extend the surveillance interval

required for performing a qualitative smoke or air flow test on the

containment spray headers. This surveillance test is not designed to

track degradation of equipment by monitoring or trending

performance. The air and smoke flow test is a test of the passive

design of the containment spray nozzles, i.e., the testing

demonstrates whether or not the nozzles are clogged. A single

failure rendering a significant number of nozzles inoperable as a

result of clogging is considered not credible. The changes being

proposed do not affect assumptions contained in plant safety

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

they affect Technical Specifications that preserve safety analysis

assumptions. Therefore, operation of the facility in accordance with

the proposed amendments would not involve a significant increase in

the probability or consequences of an accident previously analyzed.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendments extend the surveillance interval

required for performing a qualitative smoke or air flow test on the

containment spray headers. The changes being proposed will not

change the physical plant or the modes of plant operation defined in

the Facility License. The change does not involve the addition or

modification of equipment nor does it alter the design or operation

of plant systems. Therefore, operation of the facility in accordance

with the proposed amendments would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

amendments would not involve a significant reduction in a margin of

safety.

The revised surveillance interval proposed by this submittal

will not change or otherwise influence the degree of operability

assumed for the containment spray system in the plant safety

analyses. The changes being proposed do not alter the bases for

assurance that safety-related activities are performed correctly or

the basis for any Technical Specification that is related to the

establishment of or maintenance of a safety margin. Therefore,

operation of the facility in accordance with the proposed amendments

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

NRC Project Director: Herbert N. Berkow.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: April 15, 1994.

Description of amendment request: The proposed amendment requests

the deletion of the audit program frequency requirements from Technical

Specification (TS) 6.5.3 and to utilize the Operational Quality

Assurance (OQA) Plan as the controlling document. This change will

introduce more flexibility into audit scheduling to consider plant

activities and performance. In addition, a minor editorial change has

been incorporated correcting a reference in TS 6.5.1.14 in response to

a finding in the Operational Safety Team Inspection report of December

23, 1993.

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:

GPU Nuclear has determined that this [Technical Specification

change request] TSCR poses no significant hazard as defined by the NRC

in 10 CFR 50.92.

1. These changes do not affect the function of any system or

component. Therefore, they do not increase the probability of

occurrence or consequence of an accident previously evaluated in the

[Safety Analysis Report] SAR.

2. These changes do not involve a physical change to plant

configuration and they do not affect the performance of any

equipment. Therefore, they do not create the possibility of an

accident or malfunction of a different type than previously

identified.

3. The shifting of the audit frequency requirements from the

Technical Specifications to the OQA Plan and the extension of the

maximum interval between audits of certain areas do not change the

activities to be audited nor the scope of individual audits.

Furthermore, audit frequencies are not associated with the margin of

safety in the bases of any Technical Specification.

Therefore, the margin of safety is not affected by this change.

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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753.

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

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

NRC Project Director: John F. Stolz.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: April 19, 1994.

Description of amendment request: The proposed change updates and

clarifies Technical Specification 3.4.B.1 to be consistent with

existing Specifications 1.39 and 4.3.D (ASME Code Section XI, Article

5000 requirements).

The requested change would delete reference to the ASME Code

Section XI, IS-5000 ten year hydrotest inspection interval and replace

this with references to: (1) The Technical Specification 1.39

definition for Reactor Vessel Pressure Testing, and (2) the Technical

Specification 3.3.A.(i) Reactor Vessel Pressure Testing limits (P/T and

250 deg.F maximum test temperature).

The requested change will clarify that the five electromatic relief

valves' (EMRV) pressure relief function may be inoperable or bypassed

during system pressure testing required by ASME Code Section XI,

Article IWA-5000, including system leakage and hydrostatic test, with

reactor vessel completely solid, core not critical and Technical

Specification 3.2.A (Core Reactivity limits) satisfied.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The requested change will not involve a significant increase

in the probability or consequence of any accident previously

evaluated because this change: (a) Merely updates and clarifies

Technical Specification 3.4.B.1 to be consistent with other existing

Technical Specifications, (b) contains no adverse changes to any

existing safety function necessary for the reactor vessel solid,

core not critical condition, and (c) makes no modification or

physical changes to plant equipment, performance or operation

necessary to respond to accidents for the reactor vessel solid, core

not critical condition.

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

or different accident from any accident previously evaluated because

this change: (a) Merely updates and clarifies Technical

Specification 3.4.B.1 to be consistent with other existing Technical

Specifications, (b) contains no adverse changes to any existing

safety function necessary for the reactor vessel solid, core not

critical condition, and (c) over pressure protection would continue

to be provided by the code safety valves when the EMRV pressure

relief function is bypassed.

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

because even though the EMRV pressure relief function is bypassed,

over pressure protection would continue to be provided by the code

safety valves. Elimination of this relief function does not affect

the reactor safety analysis, since credit was not taken for the EMRV

pressure relief function . . . .

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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753.

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

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

NRC Project Director: John F. Stolz.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of amendment request: March 15, 1994.

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) by removing TS 3/4.3.8,

``Turbine Overspeed Protection System,'' from the TS and relocating it

to an administratively controlled document.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This change request proposes deletion of Technical Specification

3/4.3.8, ``Turbine Overspeed Protection System'' and relocates this

requirement to an existing plant program. The purpose of overspeed

protection is to minimize the possible generation of turbine

fragment missiles. Excessive overspeed could potentially result in

the generation of missiles which could impact and damage safety

related components, equipment or structures, depending on the size

and trajectory of the missiles. The proposed deletion of this

specification is based on the low probability of the generation of a

damaging turbine missile and other existing performance

verifications of the overspeed protection system.

The turbine-generator orientation at RBS [River Bend Station] is

a ``favorable'' orientation for reducing the probability of damage

to safety-related equipment from turbine missiles since all safety-

related components and structures are located in the axial direction

from the turbine-generator. Turbine Overspeed Protection System is

necessary for protection of the turbine from only an operational and

economic point of view. The system is not essential to mitigating

the consequences of an accident. The system is not used in an

initial condition of a design basis accident or transient analysis.

The probability of damage to safety-related equipment based on

turbine manufacturer's turbine failure data was calculated to be

1.473 x 10-8 per year and is acceptably low based on the

probability of turbine failure data of 4.75x10-7 per year as

recommended by NUREG-0800. Therefore, this proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The change proposes to relocate this requirement to an existing

plant program, whereby adequate control of information is

maintained. The proposed change does not necessitate a physical

alteration of the plant (no new or different type of equipment will

be installed) or changes to parameters governing normal plant

operation. The proposed change will not impose any different

operational or surveillance requirements. No new failure modes are

introduced. Therefore, this proposed change does not create the

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

previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed change will not reduce a margin of safety because

it has no impact on any safety analysis assumption. The proposed

change does not alter the scope of equipment currently required to

be OPERABLE or subject to surveillance testing, nor does the

proposed change affect any instrument setpoints or equipment safety

functions. The favorable orientation of the turbine provides a

margin of safety such that the possibility of missile damage to

safety-related equipment is acceptably low. Therefore the change

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

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

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

1400 L Street, NW., Washington, D.C. 20005.

NRC Project Director: William D. Beckner.

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

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

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

County, Texas

Date of amendment request: April 28, 1994.

Description of amendment request: The licensee proposes to revise

Technical Specification Surveillance Requirement 4.6.1.3.e to add an

option which will allow the personnel airlock pneumatic system leak

test to be completed in 8 hours with a pressure drop of 0.50 psi. The

technical specifications currently require that the door seal pneumatic

system be demonstrated operable by verifying that the system pressure

does not decay more than 1.5 psi within 24 hours. The change to an 8-

hour test will expedite return to power following an outage since the

test is on the critical path for restart following outages.

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 door pneumatic seal system pressure drop test is not altered

except for providing an option to utilize a reduced test duration. A

conservative acceptance criteria of 0.50 psi will be assigned to the

optional short duration test thus maintaining the operability of the

pneumatic seal system. The proposed change does not alter equipment

or assumptions made in previously evaluated accidents, therefore the

consequences of previously evaluated accidents are not increased.

The probability of an accident is also unaffected because the seals

are not a potential accident initiator.

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

or different kind of accident from any accident previously

evaluated.

With a conservative acceptance criteria of 0.50 psi assigned to

the optional 8 hour door pneumatic seal system pressure drop test

the capability of the door pneumatic seal system to maintain 65 psig

to the airlock seals, for a minimum of 15 days upon a loss of

instrument air, is assured. Loss of plant supply air is the accident

evaluated in the UFSAR [Updated Final Safety Analysis Report]

section 3.8.2.1.2 and plant specification 2C269SS0006. The proposed

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

accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

To ensure the pneumatic seal system pressure drop test is not

compromised, a conservative acceptance criteria of 0.50 psi will be

assigned to the 8 hour test. With the conservative acceptance

criteria, the proposed change does not involve a significant

reduction in the margin of safety previously evaluated.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

P.C., 1615 L Street, N.W., Washington, D.C. 20036.

NRC Project Director: Suzanne C. Black.

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

Station, Unit No. 1 Rockingham, New Hampshire

Date of amendment request: January 14, 1994.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) to specify the composition of

the Station Operation Review Committee (SORC) based on experience and

expertise vice organizational position, to implement a Station

Qualified Reviewer Program (SQRP), to delete the requirement for

periodic procedure reviews, to revise the time within which the Nuclear

Safety Audit Review Committee (NSARC) must issue reports and minutes,

and to incorporate a number of editorial changes. The editorial changes

would delete certain items that are no longer applicable, would remove

inconsistencies involving the names of systems and equipment and NSARC

function, composition, and use of alternates, and would correct the

value for the reactor coolant system volume. Other editorial changes

would be made for document format consistency. The proposed amendment

would affect the following TS Sections and tables: 1.31, 3.3.3.6,

3.4.1.2, 4.6.3.2, 3.7.1.2, 3/4 10.6, 5.4.2, 6.3, and 6.4, and Table

4.3-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

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

probability or consequences of an accident previously evaluated (10 CFR

50.92(c)(1)).

The proposed redefinition of the composition of the SORC would not

diminish the effectiveness of the SORC and would continue to ensure

that the SORC has the desired experience and expertise to advise the

Station Manager on all matters related to nuclear safety. The proposed

change would permit operational flexibility and eliminate the need for

an amendment whenever organizational changes occur. The proposed SQRP

would not reduce the level of procedure review, since the SORC

continues to retain responsibility to review any document requiring an

evaluation pursuant to 10 CFR 50.59. The SQRP would be limited to

reviewing procedures that do not affect nuclear safety.

Deleting the requirement to periodically review procedures would

not diminish the review process for procedures since other programmatic

requirements would continue to assure procedures are reviewed and

revised when necessary.

The proposed extension of time for preparing and forwarding NSARC

meeting minutes would not affect safe operation of the facility.

Significant safety concerns or unreviewed safety questions would still

be brought to the attention of the Senior Vice President without

waiting for the release of the NSARC meeting minutes. The change would

not impede in any manner prompt communication of significant concerns

to the Senior Vice President. The proposed changes do not affect the

manner by which the facility is operated and do not change any facility

design feature or equipment. The proposed changes involve

administrative or programmatic requirements or merely involve editorial

changes, corrections, or clarifications. Since there is no change to

the facility or operating procedures, there is no effect upon the

probability or consequences of any accident previously analyzed.

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

kind of accident from any accident previously evaluated (10 CFR

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

facility is operated and do not change any facility design feature or

equipment which affects the operational characteristics of the

facility. The proposed changes involve administrative or programmatic

requirements or merely involve editorial changes, corrections, or

clarifications.

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

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

affect the manner by which the facility is operated or involve

equipment or features which affect the operational characteristics of

the facility.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Exeter Public Library, 47

Front Street, Exeter, New Hampshire 03833.

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

International Place, Boston, Massachusetts 02110-2624.

NRC Project Director: John F. Stolz.

Northeast Nuclear Energy Company, et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: February 10, 1992, as supplemented April

14, 1994.

Description of amendment request: The proposed amendment would

remove two tables from the Technical Specifications (TS) which list

reactor trip system (RTS) instrumentation response times and engineered

safety features actuation system (ESFAS) instrumentation response

times. These tables will be placed in the Millstone 3 Technical

Requirements Manual.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant hazards

consideration because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to remove the RTS and ESFAS response times

from the Technical Specifications will not affect the operation of

the RTS and ESFAS. Operability and surveillance requirements are

still maintained in the Technical Specifications and the response

times will be included and maintained in the plant operating

procedures. A safety evaluation and PORC [Plant Operations Review

Committee] review will be required for the limits to be changed.

Since the systems will not be affected by the proposed changes,

there is no impact on the performance of these systems or the

consequences of an accident previously analyzed.

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

from any previously evaluated.

There are no new failure modes associated with the proposed

changes. Since the plant will continue to operate as designed, the

proposed changes will not modify the plant response to the point

where it can be considered a new accident.

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

The proposed changes do not have any adverse impact on the

protective boundaries nor do they affect the consequences of any

accident previously analyzed. The Technical Specification

operability and surveillance requirements will still ensure that the

systems are tested and within the limits. Changing the limits

requires a safety evaluation and PORC review which will ensure that

the licensing basis is maintained. Therefore, the proposed changes

will not impact the margin of safety as defined in the basis of any

Technical Specification.

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, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: April 22, 1994.

Description of amendment request: The proposed amendment would

delete the requirements regarding the condenser air ejector monitor

from Tables 3.3-12 and 4.3-12 of the Millstone Unit 2 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:

The proposed technical specification change has been reviewed

against the criteria of 10 CFR 50.92, and it has been determined not to

involve a significant hazards consideration (SHC). Specifically, the

proposed change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Deleting the operability and surveillance requirements for the

condenser air ejector monitor from Tables 3.3-12 and 4.3-12 of the

Millstone Unit No. 2 Technical Specifications would leave the steam

generator blowdown monitor as the primary method of monitoring and

isolating steam generator blowdown. The proposed license amendment

imposes stricter limitations on the operation of Millstone Unit No.

2, because it requires the use of a single monitor, the steam

generator blowdown monitor, to meet the requirements of Millstone

Unit No. 2 Technical Specification 3.3.3.9 (Table 3.3-12).

While NNECO [Northeast Nuclear Energy Company] is proposing to

delete the operability and surveillance requirements for the

condenser air ejector monitor from the Millstone Unit No. 2

Technical Specifications, there are no plans to change any of the

design features or functions or the condenser air ejector monitor,

or any of the specified surveillances or frequency for such

surveillances. The condenser air ejector monitor will continue to

isolate blowdown upon a high radiation alarm.

Additionally, steam generator blowdown isolation is required to

ensure compliance with 10 CFR 20. It is not required to ensure

compliance with 10 CFR 100. Therefore, the condenser air ejector

monitor does not perform any safety function. The condenser air

ejector monitor is not safety related. It is not credited in any

radiological consequence calculations presented in the Millstone

Unit No. 2 FSAR [Final Safety Analysis Report].

Based on the above, this proposed license amendment 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

form any accident previously evaluated.

The proposed license amendment does not involve any physical

changes to plant equipment or any changes to plant procedures that

would be a precursor to an accident. NNECO has no plans to change

any of the specified surveillances or frequency for such

surveillances. The condenser air ejector monitor will continue to

isolate blowdown upon a high radiation alarm. Also, the proposed

license amendment imposes stricter limitations on the operation of

Millstone Unit No. 2 because it requires the use of a single

monitor, the steam generator blowdown monitor, to meet the

requirements of Millstone Unit No. 2 Technical Specification 3.3.3.9

(Table 3.3-12). Therefore, this proposed license amendment does not

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

any accident previously evaluated.