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

NUCLEAR REGULATORY COMMISSION

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations; Biweekly Notice

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 January 20, 1995, through February 3, 1995.

The last biweekly notice was published on February 1, 1995 (60 FR

6296).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By March 17, 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 [[Page 8742]] bases of the contention and a

concise statement of the alleged facts or expert opinion which support

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

the contention at the hearing. The petitioner must also provide

references to those specific sources and documents of which the

petitioner is aware and on which the petitioner intends to rely to

establish those facts or expert opinion. Petitioner must provide

sufficient information to show that a genuine dispute exists with the

applicant on a material issue of law or fact. Contentions shall be

limited to matters within the scope of the amendment under

consideration. The contention must be one which, if proven, would

entitle the petitioner to relief. A petitioner who fails to file such a

supplement which satisfies these requirements with respect to at least

one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

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

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Carolina

Date of amendment request: January 19, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirement 4.0.3 and

its associated bases to provide for a delay period of up to 24 hours in

which to perform a surveillance which has been discovered not to have

been performed within its specified frequency. This change would adopt

the requirements of NUREG-1431, ``Standard Technical Specifications,

Westinghouse Plants.''

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.

The proposed change will reduce the requirement to unnecessarily

manipulate and challenge plant systems and equipment. The most probable

result of performing a surveillance during the delay period will be to

verify its conformance with Technical Specification requirements. Since

this change does not affect plant design, operation, or the manner in

which testing is performed, the consequences of accident scenarios

postulated in the Final Safety Analysis Report will not increase.

Therefore, 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.

The proposed change does not introduce any new equipment, nor does

it require existing systems to perform a different type of function

than they are currently designed to perform. Therefore, the proposed

change does 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.

The margin of safety is neither described or prescribed for this

specification. The proposed change simply provides additional time to

perform a surveillance and verify that the operability of equipment is

in conformance with the Technical Specification requirements.

Therefore, the proposed change does not involve a significant reduction

in [the] margin of safety.

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

this review, it appears that the 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: William H. Bateman.

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: December 23, 1994.

Description of amendment request: The proposed amendments would

increase the allowable enrichment of new fuel stored in the new fuel

storage vault (NFSV), revise the enrichment description of fuel in the

reactor core, and include references to documents previously approved

by the staff in the [[Page 8743]] Technical Specifications that provide

analytical methods used to determine core operating limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability of occurrence or consequences of any accident

previously evaluated.

The Updated Final Safety Analysis Report (UFSAR) does not consider

any accidents involving the NFSV. The Fuel Handling Accidents that are

analyzed (Section 15.7.4) include dropping of a spent fuel assembly

onto the spent fuel pool floor and breaking of all fuel rods, and

dropping of a fuel assembly inside containment onto the top of the

core.

The proposed change to increase the NFSV fuel enrichment limit from

4.0 to 4.65 weight percent U-235 does not affect any of the initiators

or precursors of any accident previously evaluated. The proposed change

will not increase the likelihood that a transient initiating event will

occur because transients are initiated by equipment malfunction and/or

catastrophic system failure. Since the proposed change does not involve

the introduction of new or redesigned plant equipment, failure

mechanisms are not affected. As a result, the probability of occurrence

of accidents previously evaluated is not significantly increased.

A new criticality analysis for the proposed change to increase the

NFSV fuel enrichment limit from 4.0 to 4.65 weight percent U-235 was

performed for the NFSV. It was determined that even in worst case

conditions the acceptance criteria was met since the maximum Keff

was determined to be well below the 0.95 limit with a 95/95

probability/confidence level. The consequences of any accident,

including a fuel handling accident involving the NFSV, are not

significantly increased.

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

or different kind of accident from any previously analyzed.

The proposed change to the Technical Specifications does 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 analysis. 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.

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

in a margin of safety.

Plant safety margins are established through LCOs, limiting safety

system settings, and safety limits specified in the Technical

Specifications. There will be no changes to either the physical design

of the plant or to any of these settings and limits as a result of

increasing the NFSV fuel enrichment limit. The change does not involve

a significant increase in the probability of occurrence or consequences

of any accident previously evaluated or create the possibility of a new

or different kind of accident from any previously analyzed.

Additionally, the revised criticality analysis demonstrates that the

maximum Keff under all postulated conditions remains below the

acceptance value of 0.95. Therefore, the change will not result in a

significant reduction in a margin of safety.

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

the probability of occurrence or consequences of any accident

previously evaluated.

The proposed change to increase the reactor core fuel enrichment

range discussed in the Design Features section of Technical

Specifications from ``between 2.2 to 4.0'' to ``up to 4.65'' weight

percent U-235 is administrative in nature and does not affect any of

the initiators or precursors of any accident previously evaluated. The

proposed change will not increase the likelihood that a transient

initiating event will occur because transients are initiated by

equipment malfunction and/or catastrophic system failure. Since the

proposed change does not involve the introduction of new or redesigned

plant equipment, failure mechanisms are not affected. As a result, the

probability of occurrence of accidents previously evaluated is not

significantly increased.

The fuel enrichment limit of each core is determined by the core

specific design and is determined to be acceptable with respect to the

accident analysis by the reload analysis and is not impacted by the

value specified in the description in the Design Features section of

Technical Specifications. This value is only provided as the highest

expected core fuel enrichment in the Design Features section discussion

of the reactor core. This change is administrative in nature and does

not affect the consequences of any accident previously evaluated.

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

or different kind of accident from any previously analyzed.

The proposed change in the reactor core fuel enrichment description

contained in the Design Features section of Technical Specifications

does 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 analysis. No safety related equipment or

function will be altered as a result of the proposed change. Also, the

procedures governing normal plant operation and recovery from an

accident are not changed by the proposed Technical Specification

change. Since no new failure modes or mechanisms are added by the

proposed change, the possibility of a new or different kind of accident

is not created.

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

in a margin of safety.

Plant safety margins are established through LCOs, limiting safety

system settings, and safety limits specified in the Technical

Specifications. There will be no changes to either the physical design

of the plant or to any of these settings and limits as a result of

increasing reactor core fuel enrichment value given in the Design

Features section of Technical Specifications. The change does not

involve a significant increase in the probability of occurrence or

consequences of any accident previously evaluated or create the

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

analyzed.

Based on the above discussion, the ability to safely shutdown the

operating unit and mitigate the consequences of all accidents

previously evaluated will be maintained. Therefore, the margin of

safety is not significantly affected.

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

the probability of occurrence or consequences of any accident

previously evaluated.

The proposed change to add three documents to the list of documents

that provide the analytical methods to determine core operating limits

is administrative in nature and does not affect any of the initiators

or precursors of any accident previously evaluated. The proposed change

will not increase the likelihood that a transient initiating event will

occur because transients are initiated by equipment malfunction

[[Page 8744]] and/or catastrophic system failure. Since the proposed

change does not involve the introduction of new or redesigned plant

equipment, failure mechanisms are not affected.

The documents have been previously reviewed and approved by the NRC

and it was determined that they provide an acceptable means to

determine core operating limits. As a result, the probability of

occurrence of accidents previously evaluated is not significantly

increased. Since the documents provide NRC approved methodologies for

determining core operating limits, the addition of the documents to

Technical Specifications or use of the documents to determine core

operating limits will not significantly increase the consequences of

any accident previously evaluated.

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

or different kind of accident from any previously analyzed.

The proposed change to add three documents to the list of documents

that provide the analytical methods to determine core operating limits

is administrative in nature and does 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 analysis. 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.

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

in a margin of safety.

Plant safety margins are established through LCOs, limiting safety

system settings, and safety limits specified in the Technical

Specifications. There will be no changes to either the physical design

of the plant or to any of these settings and limits as a result of

adding references to the new documents. The ability to 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.

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: September 19, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 4.4.A.3 to reference the

testing frequency requirements of 10 CFR Part 50, Appendix J, and to

state that NRC approved exemptions to the applicable regulatory

requirements are permitted. This proposed administrative revision

simply deletes the paraphrased language and directly references

Appendix J.

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 will provide a one-time exemption from the 10

CFR [Part] 50, Appendix J Section III.D.1.(a) leak rate test schedule

requirement. This change will allow for a one-time test interval for

Type A Integrated Leak Rate Tests (ILRTs) of approximately 70 months.

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

therefore this proposed 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. The history at IP-2 [Indian Point 2]

demonstrates that Type B and C Local Leak Rate Tests (LLRTs) have

consistently detected any excessive local leakages.

Administrative controls govern the 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, an

LLRT is performed to ensure acceptable leakage levels, following any

LLRT 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 IP-2 has been in operation it has never failed a

Type A ILRT. Therefore, a one-time exemption increasing the interval

for performing an ILRT should not result in a significant decrease in

the confidence in the leak tightness of the containment structure.

The proposed change also revises Technical Specification 4.4.A.3 to

reference the testing frequency requirements of 10 CFR [Part] 50,

Appendix J, and to state that NRC approved exemptions to the applicable

regulatory requirements are permitted. The current language of TS

4.4.A.3 paraphrases the requirements of Section III.D.1.(a) 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.1.(a) will require a submittal

from Consolidated Edison under 10 CFR 50.12 and subsequent review and

approval by the NRC prior to implementation. The proposed change is

stated generically to avoid the need for further TS changes if

different exemptions are approved in the future.

The proposed change, in itself, does not affect reactor operations

or accident analysis and has no radiological consequences. The change

provides clarification so that future Technical Specifications changes

will not be necessary to correspond to applicable NRC approved

exemptions from the requirements of Appendix J.

Therefore, the proposed change 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 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

[[Page 8745]] approximately 70 months for the ILRT. Given the test

history of IP-2 of 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 provides

clarification to a specification that paraphrases a codified

requirement.

Since the proposed change would not change the design,

configuration or method of operation of the plant, it 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 ILRTs is to ensure that

the release of radioactive materials will be restricted to those leak

paths and leak rates assumed in accident analyses. The relaxed schedule

for ILRTs does not allow for relaxation of Type B and C LLRTs.

Therefore, methods for detecting local containment leak paths and leak

rates are unaffected by this proposed change. Given that the test

history for ILRTs shows no failure during plant life, a one-time

increase of the test interval does not lead to a significant

probability of creating a new leakage path or increased leakage rates,

and the margin of safety inherent in existing accident analyses is

maintained.

The proposed Technical Specification change is administrative and

clarifies the relationship between the requirements of TS 4.4.A.3,

Appendix J and any approved exemptions to Appendix J. It does not, in

itself, change a safety limit, an LCO [limiting condition for

operation], or a surveillance requirement on equipment required to

operate the plant. The NRC will directly approve any proposed change or

exemption to [Section] III.D.1.(a) of Appendix J prior to

implementation.

Therefore, this change does not involve a significant reduction in

the margin of safety.

Based on the Safety Analysis, it is concluded that: (1) The

proposed change does not constitute a significant hazards consideration

as defined by 10 CFR 50.92 and (2) there is reasonable assurance that

the health and safety of the public will not be endangered by the

proposed change. Moreover, because this action does not involve a

significant hazards consideration, it will also not result in a

condition which significantly alters the impact of the station on the

environment as described in the NRC Final Environmental Statement.

Although the licensee has included an evaluation of a proposed

exemption to 10 CFR part 50, Appendix J requirements in the above

determination of no significant hazards consideration, only the part

related to the amendment is pertinent to this notice of proposed

amendment. The exemption request will be considered as a separate

matter on its own merits. 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: 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: October 31, 1994

Description of amendment request: The requested amendments would

remove the stroke times for the steam generator power operated relief

valves (PORVs) from Technical Specification (TS) Tables 3.6-2a and 3.6-

2b. The PORVs are part of the main steam vent to atmosphere system. The

PORV actuators have difficulty developing enough closing thrust to

adequately overcome all of the friction loads within the valves;

therefore, difficulty exists in consistently meeting the present 5-

second closing stroke time requirement. The licensee requests the

proposed change on the basis that the PORVs do not receive an actual

containment isolation signal; therefore, it is justified to remove the

stroke times from TS Tables 3.6-2a and 3.6-2b.

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 48 FR 14870, the Commission has set forth examples of amendments

that are considered not likely to involve significant hazards

considerations. Example (vi) describes a change which either may result

in some increase to the probability or consequences of a previously-

analyzed accident or may reduce in some way a safety margin, but where

the results of the change are clearly within all acceptable criteria

with respect to the system or component specified in the Standard

Review Plan. In this case, the proposed amendment is similar to example

(vi) in that it removes the required isolation time of the steam

generator PORVs from TS Tables 3.6-2a and 3.6-2b; however, no adverse

impact upon accident analyses is created as a result.

Criterion 1

The requested amendments will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The effects of the delays in isolation times on the various transients

affected have been analyzed and found to be acceptable. Since these

valves do not receive a containment isolation signal, and no credit is

taken for operation of these valves in the dose analysis for a

containment isolation function, a maximum stroke time does not apply

for containment isolation.

Criterion 2

The requested amendments will not create the possibility of a new

or different kind of accident from any accident previously evaluated.

SV PORV closure (provided the valves are not already closed at the

start of the transient) is a response to a transient already in

progress. The possibility of a spurious SV PORV opening will not be

affected by the requested amendments. No equipment or component

reconfiguration will occur as a result of this change. Finally, no

changes to plant procedures are being made which would affect any

accident causal mechanisms.

Criterion 3

The requested amendments will not involve a significant reduction

in a margin of safety. The isolation times which are applicable to

these valves are specified in TS Table 3.3-5, Engineered Safety

Features Response Times. The effects of the isolation of these valves

were evaluated based on their ESF function, not a containment isolation

function, and determined to be acceptable.

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

[[Page 8746]] 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: Herbert N. Berkow

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

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

Date of amendment request: June 13, 1994, as supplemented August

15, 1994.

Description of amendment request: The proposed changes would

increase the initial fuel enrichment limit from a current maximum of

4.0 weight % to 4.75 weight % and establish new loading patterns for

new and irradiated fuel in the spent fuel pool to accommodate this

increase. These changes would also increase the efficiency of fuel

storage cell use in the spent fuel pools and provide additional

flexibility to the reload design efforts at Duke Power Company, while

at the same time maintaining sufficient criticality safety margin and

decay heat removal capabilities.

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.

There is no increase in the probability or consequences of an

accident in the new fuel vault since the only credible accidents for

this area are criticality accidents and it has been shown that

calculated, worst case Keff for this area is 0.95 under

all conditions.

There is no increase in the probability of a fuel drop accident in

the Spent Fuel Storage Pool since the mass of an assembly will not be

affected by the increase in fuel enrichment. The likelihood of other

accidents, previously evaluated and described in Section 9.1.2 of the

FSAR [Final Safety Analysis Report], is also not affected by the

proposed changes. In fact, it could be postulated that since the

increase in fuel enrichment will allow for extended fuel cycles, there

will be a decrease in fuel movement and the probability of an accident

may likewise be decreased. There is also no increase in the

consequences of a fuel drop accident in the Spent Fuel Pool since the

fission product inventory of individual fuel assemblies will not change

significantly as a result of increased initial enrichment. In addition,

no change to safety related systems is being made. Therefore, the

consequences of a fuel rupture accident remain unchanged. Also, it has

been shown that keff is 0.95, under all conditions

therefore, the consequences of a criticality accident remain unchanged

as well.

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 do not create the possibility of a new or

different kind of accident since fuel handling accidents (fuel drop and

misplacement) are not new or different kinds of accidents. Fuel

handling accidents are already discussed in the FSAR for fuel with

enrichments up to 4.1 weight %. As described in Section VI.9 of

Attachment IV, additional analyses have been performed for fuel with

enrichment up to 4.75 weight %. Worst case misloading accidents

associated with the new loading patterns were evaluated. For all

possible misloading accidents the negative reactivity provided by

soluble boron maintains keff 0.95. of safety.

3. The proposed changes do not involve a significant reduction in

the margin of safety.

The proposed change does not involve a significant reduction in the

margin of safety since, in all cases, a keff 0.95 is

being maintained. Criticality analyses have been performed which show

that the new fuel storage vault will remain subcritical under a variety

of moderation conditions, from fully flooded to optimum moderation. As

discussed above, the Spent Fuel Pool will remain sufficiently

subcritical during any fuel misplacement accident.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

Date of amendment request: November 11, 1994, as supplemented

January 30, 1995.

Description of amendment request: The amendments would revise the

Technical Specifications Design Features section to establish

restricted loading patterns and associated burnup criteria for placing

fuel in the Oconee Spent Fuel Pools. These changes are necessary to

address two new fuel designs which have increased initial fuel

enrichment and therefore cannot be stored in the spent fuel pools under

existing 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:

Standard 1. The proposed amendments will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

Each accident analysis addressed in the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to changes in

Cycle 15 parameters to determine the effect of the Cycle 16 reload and

to ensure that the acceptance criteria of the FSAR safety analyses

remain satisfied. The transient evaluation of Cycle 16 is considered to

be bounded by previously accepted analyses. Section 7 of the Reload

Report addresses ``Accident and Transient Analysis'' for this core

reload.

There is no increase in the probability or consequences of an

accident due to the spent fuel storage restrictions proposed in this

amendment request. It has been shown that the calculated, worst case

keff for this area is [less than or equal to] 0.95 under all

conditions. There is no increase in the probability of a fuel drop

accident in the SFP [spent fuel pool] since the mass of the new

assemblies is not significantly different from the mass of the old

assemblies. The likelihood of other accidents, previously evaluated and

described in the FSAR, is also not affected by the proposed changes. In

fact, it could be postulated that since the increase in fuel enrichment

will allow for extended fuel cycle lengths, there will be a decrease in

fuel movement and the probability of an accident may actually be

reduced. There is also no increase in the consequences of a fuel rod

drop accident in the SFP since the fission product inventory of

[[Page 8747]] individual fuel assemblies will not change significantly

as a result of increasing the initial enrichment. In addition, no

change to safety related systems is being made. Therefore, the

consequences of a fuel rupture accident remain unchanged. In addition,

it has been shown that keff is [less than or equal to] 0.95 under

all conditions. Therefore, the consequences of a criticality accident

in the SFP remain unchanged as well. The above analysis ensures that

the proposed reload amendment request 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.

The analyses performed in support of this reload are in accordance

with the NRC approved methods delineated in Specification 6.9.2. The

predicted operating characteristics of Oconee 3 Cycle 16 are similar to

previously licensed designs. The Mark B10T and Mark B11 fuel assembly

designs remain mechanically compatible with all fuel handling

equipment. Therefore, no new or different kind of fuel handling

accident is created by the proposed amendment request.

Section 15.11 of the Oconee FSAR states that the refueling boron

concentration is maintained such that a criticality accident during

refueling is not considered credible. The proposed amendment request

continues to assure that a criticality accident in the SFP or during

refueling is not credible. The double contingency principle discussed

in ANSI N-16.1-1975 and the April 1978 NRC letter allows credit for

soluble boron under other abnormal or accident conditions, since only a

single accident need be considered at one time. Thus, by requiring a

minimum boron concentration in the SFP, a criticality accident caused

by violating the SFP storage restrictions is not considered credible.

Therefore, the proposed amendment request does 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

the margin of safety.

The Oconee 3 Cycle 16 design was performed using the NRC approved

methods given in Specification 6.9.2. The safety limits for Oconee 3

Cycle 16 are unchanged from previous cycles. The limits and margins

summarized in the Oconee 3 Cycle 16 Reload Report are well within the

allowable limits and requirements, and reflect no reductions to any

margins of safety.

The proposed change does not involve a significant reduction in the

margin of safety related to SFP criticality. In all cases, a keff

[less than or equal to] 0.95 is maintained. Criticality analyses have

been performed which show that the SFP will remain sufficiently

subcritical during any fuel misplacement accident. In summary the

proposed changes do not involve a significant reduction in the margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036.

NRC Project Director: Herbert N. Berkow.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: January 6, 1995

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) 3/4.8.1.1, ``AC Sources-

Operating,'' and 3/4.8.1.2, ``AC Sources-Shutdown,'' to (1) revise the

minimum quantity of fuel oil required in the day tanks and the storage

tanks, (2) add specific actions to be taken if the storage tank levels

fall below minimum requirements, (3) revise and relocate to the

associated Bases the fuel oil sampling and testing criteria, and (4)

add specific actions to be taken if the fuel oil properties do not meet

specified limits. The proposed amendment would also revise TS 6.8.4,

``Programs,'' to add a requirement for a diesel fuel oil testing

program. The licensee stated that the proposed changes are consistent

with the NRC's Improved Standard Technical Specifications (NUREG-1434).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The diesel generators are not initiators or precursors to an

accident previously evaluated. The diesel generators are required to

provide onsite power to safe shutdown loads as assumed in the accident

analysis. Therefore, the proposed changes to the diesel generator fuel

oil specifications cannot significantly affect the probability of a

previously evaluated accident.

The proposed change to the minimum required diesel generator fuel

oil levels is based on updated calculations of fuel consumption rates.

Because the updated calculations assume a lower consumption rate, the

new minimum fuel oil levels are lower but still assure that a seven-day

fuel oil capacity is available. Accordingly, the proposed change has no

effect on the operation of the diesel generator. The proposed change to

allow 48-hours to restore diesel generator fuel oil to the minimum

required level does not affect short-term diesel generator operability

and is acceptable based on the remaining fuel oil capacity (>6 days),

initiating the process for procuring additional fuel and the low

probability of an event requiring a diesel generator during this

interval. Also, the proposed allowance of a limited time to restore

diesel fuel oil properties to required limits will not affect the

short-term operability of the diesel generator. Even with minor

degradation of the fuel oil properties, the diesels will start and

perform their intended function. Relocation of the testing requirements

to the bases and adding a description of the Diesel Fuel Oil Testing

Program to the Administrative Control section are administrative

changes. The diesel fuel oil will continue to be sampled and tested in

a manner to assure its quality. In summary, the changes will not

adversely affect the performance or the ability of the diesel

generators to perform their intended function. Therefore, the proposed

changes will not significantly increase the consequences of an accident

previously evaluated.

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

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

different kind of accident from any accident previously evaluated.

The proposed changes will revise the minimum required diesel

generator fuel oil levels and requirements associated with diesel

generator fuel oil properties. [[Page 8748]] The changes do not

introduce any new accident precursors and do not involve any

alterations to plant configurations which could initiate a new or

different kind of accident. The proposed changes do not affect the

short-term operability of the diesel generator. In addition, the

operability of the diesel generators is assured by periodic testing and

preventive maintenance. Therefore, the proposed changes will not create

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

accident previously evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

Safety margins are established through safety analyses. These

analyses assume that at least one diesel generator will start and load

whenever offsite power is lost. The proposed change to the minimum

required diesel generator fuel oil levels is based on updated

calculations of fuel consumption rates. The updated calculations use

the guidance delineated in Regulatory Guide 1.137 which is based on

time-dependent loads of the diesel-generators during design basis

events. Calculations based on time dependent loads result in new

minimum fuel oil levels which are lower. This change has no effect on

the operation of the diesel generator or on a margin of safety. The

allowance of a limited time to restore the fuel oil levels, or to

analyze and restore fuel oil properties to required limits, is

justified since the short term operability of the diesel generators is

not affected. Relocation of the fuel oil testing requirements to the

Bases does not affect the quality of the fuel oil. The 10CFR50.59

process will assure that future changes to the Bases will maintain the

current margins of safety, and that the diesel fuel oil will continue

to be sampled and tested in such a manner as to assure its quality.

Adding a description of the Diesel Fuel Oil Testing Program to the

Administrative Control section of Technical Specifications are

administrative. Therefore, the diesel generator will continue to

operate as analyzed and there will not be a significant reduction in a

margin of safety.

The proposed changes are further justified in that they are

consistent with the requirements of the Improved Standard Technical

Specifications (NUREG-1434).

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Ledyard B. Marsh.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: January 6, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) 3/4.3.7.5, ``Accident Monitoring

Instrumentation,'' and TS 3/4.4.2, ``Safety/Relief Valves.'' TS 3/

4.3.7.5 would be revised to delete certain instruments not classified

as Category 1 (Type A or non-Type A) as defined in Regulatory Guide

1.97 and to delete the requirement that accident monitoring

instrumentation be operable in Operational Condition 3. The ACTIONS of

TS Table 3.3.7.5-1 would be revised to allow 30 days to restore one

inoperable channel and 7 days to restore two inoperable channels. TS

3.3.7.5 would be revised to add an exception to the requirements of TS

3.0.4. In addition, editorial changes would be made to TS Tables

3.3.7.5-1 and 4.3.7.5-1 for consistency and clarity.

The proposed amendment would also revise TS 3/4.4.2 to remove

requirements related to safety/relief valve acoustic monitors to be

consistent with the proposed changes to TS Tables 3.3.7.5-1 and

4.3.7.5-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of NMP2 [Nine Mile Point Nuclear Station Unit 2] in

accordance with the proposed amendment, will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

PAM [Post-Accident Monitoring] instruments are used to help guide

operator response to postulated accidents. Thus, the status or

operability of PAM instrumentation does not affect the probability of

previously analyzed accidents. The non-Category 1 PAM instruments being

removed from the Technical Specifications do not meet any of the

Commission's screening criteria and are not of controlling importance

to safety or necessary to obviate the possibility of an abnormal

situation or event giving rise to an immediate threat to public health

and safety. The operability of critical parameters necessary to assure

proper response to previously analyzed accidents (i.e., Category 1

instruments) is still controlled by the Technical Specifications. Thus,

deleting non-Category 1 instruments will not increase the consequences

of any accident previously evaluated.

PAM instruments are related to the diagnosis and preplanned actions

required to mitigate DBAs [Design Basis Accidents] assumed to occur in

Operational Conditions 1 and 2. A DBA during Operational Condition 3 is

extremely unlikely. The requirement to maintain the Reactor Water

Level, Suppression Pool Water Level and Drywell High Range Radiation

Monitor instrumentation operable in Operational Condition 3 will be

deleted. Because Suppression Pool Water Level indication will no longer

be required in Operational Condition 3, its ACTION requirement was

revised to delete the requirement to place the plant in COLD SHUTDOWN,

Operational Condition 4. This is consistent with ITS [Improved Standard

Technical Specifications] which requires that the plant be brought to

an operational condition in which the LCO [Limiting Condition for

Operation] does not apply if a required action cannot be met.

Therefore, deleting the requirement that PAM instruments be operable

during Operational Condition 3 and changing the ACTION requirement for

Suppression Pool Water Level Monitoring does not affect the probability

or consequences of an accident.

The passive nature of the Category 1 PAM instruments (i.e., those

instruments that initiate no critical automatic action) and the

alternate means available to obtain the required information assure an

acceptable level of safety is maintained during operation with

instrument channels out of service. Since an acceptable level of safety

is maintained with inoperable channels, plant startup or operation with

inoperable channels will not alter plant response to analyzed

accidents. Thus, the proposed changes to the required ACTIONS and the

proposed exemption to Specification 3.0.4 will not increase the

consequences of analyzed events.

The proposed changes to the requirements for PCIV [Primary

[[Page 8749]] Containment Isolation Valve] indication are consistent

with the proposed required ACTIONS. Position indication will still be

required for each operable PCIV and penetrations without adequate PCIV

indication status will be isolated, thus assuring containment integrity

in the event of an accident. Deletion of the ``Minimum Required

Actions'' column in Table 3.3.7.5-1 is consistent with the proposed

ACTIONS for LCO 3.3.7.5, since compensatory actions are based on

compliance with the ``Required Number of Channels.'' Deleting the

``Applicable Operating Conditions'' column is consistent with the

proposed changes and other NMP2 Technical Specifications sections.

Finally, referencing Specification 4.0.5 is an administrative change

which does not alter any existing surveillance requirements for the

safety relief valves.

In aggregate, the proposed changes do not affect the plant in a way

that could directly contribute to causing or mitigating the effects of

an accident. Therefore, the operation of NMP2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The operation of NMP2, in accordance with the proposed amendment,

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

from any accident previously evaluated.

The proposed changes do not represent a physical change to the

plant as described in the NMP2 USAR [Updated Safety Analysis Report].

The proposed changes do not modify any plant equipment and the initial

conditions used for the design basis accident analysis are still valid.

Thus, no potential initiating events are created which would cause any

new or different kinds of accidents. PAM instrumentation is used to

guide operator response during postulated accidents. Those PAM

instruments considered of controlling importance to safety are retained

in the Technical Specifications. Thus, plant response to previously

analyzed events is not altered so as to create any new or different

kinds of accidents. Therefore, operation of Nine Mile Point Unit 2 in

accordance with the proposed change will not create the possibility of

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

The operation of NMP2, in accordance with the proposed amendment,

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

The non-Category 1 PAM instruments being removed from the Technical

Specifications do not meet any of the Commission's screening criteria.

That is, the instruments being proposed for removal are not of

controlling importance to safety or necessary to obviate the

possibility of an abnormal situation or event giving rise to an

immediate threat to public health and safety. Thus, they are not

critical to any margin of safety.

PAM instruments are related to the diagnosis and preplanned actions

required to mitigate DBAs assumed to occur in Operational Conditions 1

and 2. A DBA during Operational Condition 3 is extremely unlikely. The

requirement to maintain the Reactor Water Level, Suppression Pool Water

Level and Drywell High Range Radiation Monitor instrumentation operable

in Operational Condition 3 will be deleted. Because Suppression Pool

Water Level indication will no longer be required in Operational

Condition 3, its ACTION requirement was revised to delete the

requirement to place the plant in COLD SHUTDOWN, Operational Condition

4. This is consistent with the ITS, which requires that the plant be

brought to an operational condition in which the LCO does not apply if

a required action cannot be met. Therefore, deleting the requirement

that PAM instruments be operable during Operational Condition 3 and

changing the ACTION requirement for Suppression Pool Water Level

Monitoring does not significantly reduce a margin of safety.

Since the Category 1 PAM instruments are passive in nature (i.e.,

no critical automatic action is assumed to occur from these

instruments) and alternate means exist to obtain the required

information, an acceptable level of safety is assured when instrument

channels are out of service. Also, the probability of an event

requiring PAM instrumentation is low. Continued operation with one

channel out of service, and limited plant operation with two channels

out of service, does not compromise plant safety margins. An acceptable

level of safety is maintained during plant startups and operation with

instrument channels out of service. Thus, the proposed changes to the

required ACTIONS and the proposed exemption to Specification 3.0.4 will

not significantly reduce a margin of safety.

The proposed changes to PCIV indication will assure correct

implementation of the ACTIONS discussed above. Isolating the flow path

associated with one or two inoperable PCIV indication channels is

conservative since the subject valve will be positioned as required to

assure primary containment integrity. The remaining editorial changes

are administrative in nature and by definition do not affect safety

margins. Deleting the ``Minimum Operable Channels'' and ``Applicable

Operating Conditions'' columns is consistent with the proposed changes.

Finally, referencing the requirements of Specification 4.0.5 is an

administrative change and by definition does not reduce the margin of

safety.

Therefore, the operation of NMP2 in accordance with the proposed

change will not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Ledyard B. Marsh.

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: December 9, 1994.

Description of amendment request: The proposed changes incorporate

NRC recommendations contained in Generic Letter 93-05 related to the

diesel generator (DG) surveillance requirements and other DG

surveillance requirements related to the cold starts. The proposed

changes to the DG operability testing surveillance requirements are

consistent with the intent of GL 93-05 however vary in some

particulars, because of circumstances specific to Millstone 3. The

proposed changes will modify the requirement for the DG operability

testing when the other DG is inoperable, delete the requirement for DG

operability testing when one or both offsite AC sources are inoperable,

eliminate fast loading of DGs except for the 18-month test, and modify

the hot restart test from the 24-hour loaded test run for the DGs.

Basis for proposed no significant hazards consideration

determination: [[Page 8750]] As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the issue of no significant

hazards consideration (SHC), which is presented below:

* * * The proposed changes do not involve a SHC because the changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed changes revise the action requirements regarding

operability testing of a non-affected DG when the other DG is

inoperable, delete the requirement for operability testing of the DGs

when one or both offsite AC sources are inoperable and eliminate the

fast loading of DGs except for the 18-month test. These changes will

improve DGs performance by reducing the number of unnecessary quick

starts and by requiring more appropriate testing of the DGs when there

is a potential for common mode failure. The proposed change, to revise

the method of verifying DG hot restart capability after a 24-hour run

without loading the DG with LOP/SI [loss of offsite power/safety

injection] load, meets an intent of Regulatory Guide 1.108, Position

C.2.a.5, which states the purpose of the test as to ``demonstrate

functional capability at full load temperature conditions.'' Functional

capability of the DG can be adequately demonstrated by manually or

automatically restarting the DG within five minutes after a 24-hour

test run without loading it with LOP/SI loads, provided that a full

load temperature condition is maintained prior to restart. The proposed

DG restart method does not reduce the effectiveness of the test. The

proposed revisions of the DG surveillance requirements will not

increase the probability of an accident and it will not change the

response of the DG to a LOP as described in the Millstone Unit No. 3

FSAR. Since the plant response to an accident will not change, there is

no change in the potential for an increase in the consequences of an

accident previously analyzed.

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

previously evaluated.

The proposed changes of the DG surveillance requirements and

operability testing requirements do not affect the operation or

response of any plant equipment or introduce any new failure

mechanisms. The proposed changes do not affect the test results and the

DGs will be verified to be operable and their response to a loss of

voltage will be unchanged. The plant equipment will respond per the

design and analyses and there will not be a malfunction of a new or any

type introduced by the revision to the DG surveillance requirements. As

such, the changes do not create the possibility of a new or different

kind of accident previously evaluated.

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

The bases of Technical Specification 3/4.8, ``Electrical Power

Systems,'' state that the operability of the AC and DC power systems

and associated distribution systems ensure that sufficient power will

be available to supply the safety-related equipment required for safe

shut down and mitigation and control of accident conditions. The bases

also state that the surveillance requirements for determining the

operability of the DGs are in accordance with the recommendations of

Regulatory Guide 1.108, Revision 1. The revisions of the surveillance

requirements establishes tests that will continue to verify that the

DGs are operable and the testing will still meet the intent of

Regulatory Guide 1.108, Revision 1. Operable DGs ensure that the

assumptions in the bases of the Technical Specifications are not

affected and ensure that the margin of safety is not reduced.

Therefore, the assumptions in the bases of the technical specifications

are not affected and these changes do 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: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: Phillip F. McKee.

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: December 14, 1994.

Description of amendment request: The proposed amendment would

revise the Millstone Unit No. 3 Technical Specifications by:

1. Increasing the upper bound of the overall containment integrated

leakage rate required by Technical Specification 3.6.1.2.a from 0.3 wt.

% per day to 0.65 wt. % per day of the containment air per 24 hours at

design basis pressure.

2. Revising Technical Specification 4.6.6.1.d.3 by providing more

margin with respect to the drawdown time for secondary containment

vacuum.

3. Revising Bases Section 3/4.7.9 to reflect the above changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

* * * 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 evaluated.

* * * There is a reasonable assurance that the modified criteria for

the negative pressure in the secondary containment boundary proposed

via the proposed change (i.e., a negative pressure of 0.1 inches in one

minute and a negative pressure of 0.4 inches within the next two

minutes), can be accomplished in the prescribed time.

Extension of the time allowed to achieve the final drawdown of

secondary containment from 120 seconds to 180 seconds (these times

include the diesel generator start and load time of approximately 11

seconds) will have a negligible impact on heating and cooling. Plant

experience has shown that heatup and cooldown of thick-walled concrete

structures, such as the Millstone Unit No. 3 auxiliary building, is a

relatively slow process. Also, natural convection within the auxiliary

building tends to stabilize temperatures. Following an accident signal,

ventilation equipment is restarted promptly. Therefore, heatup or

cooldown, during short periods while ventilation fans and/or heaters

are inactive, is insignificant and can be neglected.

The proposed change to reinstate the containment integrated leakage

rate at the design basis pressure from 0.3 wt % per day to 0.65 wt %

per day has been evaluated to determine the impact to the Appendix J

requirements for Type A, B and C Testing. In addition, the radiological

consequence evaluation also addressed the increase in La (i.e.,

from 0.3 wt % per day to 0.65 wt % per day).

On October 12, 1993, Millstone Unit No. 3 successfully conducted

the second [[Page 8751]] Type A test in the first 10-year service

period. Test results indicated that the ``As-Found'' and ``As-Left''

ILRTs [integrated leakage rate tests] passed the technical

specification acceptance criteria. The ``As-Found'' value was 0.1327

weight percent per day and the ``As-Left'' value was 0.1313 weight

percent per day. These values represent 27.2% and 26.9% of the

technical specification criterion of 0.4875 wt % per day (0.75

La), based on La equal to 0.65 wt % per day, respectively. In

addition, as of October 9, 1993, the total Type B and C ``As-Found''

and ``As-Left'' leakage results were 0.099 wt % per day, and 0.084 wt %

per day, respectively. These values represent approximately 25.3% and

21.5% of the technical specification limit of 0.39% wt % per day (0.6

La), based on La equal of 0.65 wt % per day, respectively.

Correspondingly, the 1993 Type A, B, and C test results indicate that

the ``As-Found'' and ``As-Left'' result in each test case was below the

existing Technical Specification limit of 0.3 wt % per day. This

further demonstrates the overall leakage integrity of the containment

and its boundaries.

Based on the relatively low ``As-Left'' ILRT leakage rate (i.e.,

0.1313 wt % per day is well below the existing technical specification

limit of 0.225 wt % per day (0.75 La), based on La equal to

0.3 wt % per day), which represents the overall containment integrated

leakage rate for the containment prior to start-up, there is reasonable

assurance that containment integrity will be maintained below the

allowable leakage rate limit of 0.65 wt % per day. In addition, the

total Type B and C ``As-Left'' leakage result of 0.084 wt % per day

(this is well below the existing technical specification limit of 0.18

wt % per day (0.6 La), based on La equal to 0.3 wt % per

day), provides further assurance that leakage, based on individual

penetration, will be maintained within sufficient margin of the leakage

limits.

Because the last Type A, B, and C tests were performed under the

technical specification limit of 0.65 wt % per day, the proposed change

to restore La to 0.65 wt % per day has no impact to these systems

from a leakage allowance perspective. As indicated above, the previous

test results met the technical specification leakage limits (based on

0.65 wt % per day) within sufficient margin and, therefore, would not

present any challenge to these leakage limits.

NNECO has evaluated the proposed changes to Surveillance

Requirement 4.6.6.1.d.3 that increase the time to draw a final required

negative pressure as measured at the 24'-6'' elevation of the auxiliary

building in conjunction with the proposed change to reinstate the

containment integrated leakage rate of 0.65 wt % per day to determine

the impact on the offsite doses following a LOCA. The calculated

radiological doses are, in most cases, less than the previously

calculated doses (i.e., EAB [exclusion area boundary] and LPZ [low-

population zone] doses) and are within the 10CFR100 limits. Previously,

the EAB thyroid and whole body doses as documented in the November 4,

1993, submittal were calculated to be 141 REM and 9.4 REM respectively,

while the previously docketed (i.e., the November 4, 1993, submittal)

LPZ doses to the thyroid and whole body were calculated to be 29.8 REM

and 1.7 REM respectively. Utilizing the revised application of

containment recirculation spray DF, the EAB thyroid and whole body

doses were calculated to be 61 REM and 16.7 REM, respectively, and the

LPZ thyroid and whole body doses were calculated to be 10.9 REM and 2.8

REM respectively. The assumptions used in the above radiological dose

calculations are provided in Attachment 1. It is noted that a LOCA at

Millstone Unit No. 3 is also one of the bounding accidents for the

Millstone Unit No. 3 control room, Millstone Unit No. 2 control room,

and the Millstone Technical Support Center habitability analysis.

Therefore, the doses for these areas were recalculated and are

presented in the Safety Assessment section above. The Millstone Unit

No. 1 control room and the Emergency Operating Facility doses are

bounded by the Millstone Unit No. 1 LOCA calculations.

The Millstone Unit Nos. 2 and 3 control rooms and Millstone

Technical Support Center doses were not recalculated in 1993 (i.e.,

November 4, 1993, submittal) since EAB/LPZ doses proved that the

releases were less than the 1990 submittal. In summary, all control

room and Technical Support Center doses are within the guidelines of

GDC 19. Therefore, the proposed changes do not result in an increase in

consequences of an accident (i.e., a LOCA) previously analyzed.

The proposed changes to Bases Section 3/4.6.6 do not have any

safety impact since they only reflect the changes proposed to

Surveillance Requirement 4.6.6.1.d.3.

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

from any accident previously evaluated.

The proposed changes do not compromise the ability of the SLCRS

[supplementary leak collection and release system] and ABFS [auxiliary

building filter system] to mitigate the consequences of an accident.

The proposed changes do not make any physical or operational changes to

existing plant structures, systems or components. The proposed changes

do not introduce any new or unique operational modes or accident

precursors. Therefore, the proposed changes do not create the

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

previously evaluated.

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

NNECO has evaluated the proposed changes to Surveillance

Requirement 4.6.6.1.d.3 that increase the time to draw a final required

negative pressure as measured at the 24'-6'' elevation of the auxiliary

building in conjunction with the proposed change to reinstate the

containment integrated leakage rate of 0.65 wt % per day to determine

the impact on the offsite doses following a LOCA. The calculated

radiological doses are, in most cases, less than the previously

calculated doses and these doses are within the 10CFR100 limits. All

control rooms and technical support center doses are within the

guidelines of GDC 19. Therefore, the proposed changes do not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: Phillip F. McKee.

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: December 23, 1994.

Description of amendment request: The proposed amendment would

change the acceptance criteria for the peak transient generator voltage

from 4784 volts to 5000 volts during full load rejection tests of the

diesel generator (DG), and delete the 10-year surveillance requirement

to perform a [[Page 8752]] 110% pressure test of the DG fuel oil

system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

* * * The proposed changes do not involve a SHC because the changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

DG Full-Load Rejection Test

NNECO is proposing to modify Surveillance Requirement 4.8.1.1.2.g.3

of the Millstone Unit No. 3 Technical Specifications by changing the

acceptable transient voltage to 5000 volts from 4784 volts. This change

will permit the DG full load rejection tests to be performed at

realistic plant conditions using a power factor that will envelope the

calculated power factor during the worst kW loading conditions. The

transient voltage of 5000 volts is within the normal design limits of

the DGs.

The proposed change does not alter the intent of the surveillance,

does not involve any physical changes to the plant, does not alter the

way any structure, system, or component functions, and does not modify

the manner in which the plant is operated. As such, the proposed change

to Surveillance Requirement 4.8.1.1.2.g.3 will not degrade the

capability of the DGs to perform their intended safety function, and

will not reduce the availability of the DGs. Actually, the proposed

change will increase the effectiveness of the full load rejection

tests, because the DGs will be tested in a configuration that is closer

to the design basis conditions.

Pressure Test of the DG Fuel Oil System

The DG fuel oil system is classified as an ASME Code Class 3 system

in accordance with the guidance of Regulatory Guide 1.26, ``Quality

Group Classification and Standards for

Water-, Steam-, and Radioactive-waste Components of Nuclear Power

Plants.'' Surveillance Requirement 4.0.5 requires the testing of ASME

Class 1, 2, and 3 components in accordance with Section XI of the ASME

Code. Surveillance Requirement 4.8.1.1.2.i.2 is redundant to the ASME

Section XI pressure test requirements of Surveillance Requirement

4.0.5. Additionally, the DG fuel oil tank cannot be tested in the

configuration required by Surveillance Requirement 4.8.1.1.2.i.2,

because the tanks are vented to the atmosphere and the vent cannot be

isolated. Therefore, NNECO is proposing to delete Surveillance

Requirement 4.8.1.1.2.i.2.

The proposed change does not modify the manner in which the DGs

respond to an accident. Also, the proposed change does not reduce the

reliability of the DGs.

Conclusion

Based on the above, the proposed changes to Surveillance

Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No.

3 Technical Specifications do not involve a significant increase in the

probability or consequences of an accident previously analyzed.

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

from any previously analyzed.

DG Full-Load Rejection Test

The DGs are required to operate in response to a loss of offsite

power. Their failure cannot initiate an accident. Additionally, the

proposed change to Surveillance Requirement 4.8.1.1.2.g.3 does not

affect the operation or response of any plant structure, system, or

component, and it does not introduce any new failure mechanisms.

Pressure Test of the DG Fuel Oil System

The proposed change to Surveillance Requirement 4.8.1.1.2.i.2 does

not affect the design or function of the DG fuel oil system. Failure of

the DG fuel oil system would not initiate an accident.

Conclusion

Based on the above, the proposed changes to Surveillance

Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No.

3 Technical Specifications will not create the possibility of a new or

different kind of accident from any previously evaluated.

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

DG Full-Load Rejection Test

NNECO is proposing to modify Surveillance Requirement 4.8.1.1.2.g.3

of the Millstone Unit No. 3 Technical Specifications by changing the

acceptable transient voltage to 5000 volts from 4784 volts. The intent

of the proposal is to permit the DG full load rejection tests to be

conducted at conditions which simulate design basis conditions.

The proposed change does not alter the intent of the surveillance,

does not involve any physical changes to the plant, does not alter the

way any structure, system, or component functions, and does not modify

the manner in which the plant is operated. As such, the proposed change

to Surveillance Requirement 4.8.1.1.2.g.3 will not degrade the ability

of the DGs to perform their intended safety function, and will not

reduce the availability of the DGs.

The bases of Technical Specification 3/4.8, ``Electrical Power

Systems,'' state that the operability of the AC and DC power systems

and associated distribution systems ensure that sufficient power will

be available to supply the safety related equipment required for safe

shutdown and for the mitigation of transients. The proposed change to

the surveillance requirement will increase the effectiveness of the

full load rejection tests.

This will ensure the operability of the DGs. Operable DGs ensure

that the assumptions for the bases of the Millstone Unit No. 3

Technical Specifications are not affected.

Pressure Test of the DG Fuel Oil System

NNECO is proposing to delete Surveillance Requirement 4.8.1.1.2.i.2

from the Millstone Unit No. 3 Technical Specifications. This

surveillance requirement is redundant to the requirements of

Surveillance Requirement 4.0.5 which invokes ASME Section XI.

Additionally, the fuel oil system cannot be tested to the requirements

of Surveillance Requirement 4.8.1.1.2.i.2 because the DG fuel oil tanks

are vented to the atmosphere and this vent path cannot be isolated.

Millstone Unit No. 3 will include the DG fuel oil system pressure

test as an augmented inspection within the Inservice Inspection

program. Inspections will be performed in compliance with the

requirement of the 1983 Edition of ASME Section XI, Table IWD-2500-1,

``Test and Examination Categories.'' Testing (i.e., a system

hydrostatic test) in accordance with ASME Section XI will provide

equivalent assurance of tank and piping integrity.

Conclusion

Based on the above, the proposed changes to Surveillance

Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No.

3 Technical Specifications do not involve a significant reduction in

the margin of safety.

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

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

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

[[Page 8753]] amendment request involves no significant hazards

consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: Phillip F. McKee.

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: January 18, 1995.

Description of amendment request: The proposed changes to the

technical specifications will increase the minimum required boron

concentration in the boric acid tank (BAT) from 6300 ppm to 6600 ppm.

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 (SHC), which is presented below:

* * * 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 evaluated.

The change affects the minimum required boron concentration in the

BAT. Changes in the tank's boron concentration will not affect the

probability of any plant accident.

An increase in the minimum BAT concentration of 6600 ppm was

recommended by Westinghouse based on their Cycle 6 BORDER evaluation.

The BORDER evaluation conservatively determines the ability to maintain

shutdown margin when the plant is taken from an initial operating

condition of Mode 1 or 2 to a final condition of Mode 5 or 6 using an

assumed minimum BAT concentration. Therefore, the ability to maintain

shutdown margin is assured and the change will not adversely affect the

consequences of any plant accident.

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

from any Previously Analyzed.

The change conservatively increases the minimum required boron

concentration in the BAT from 6300 ppm to 6600 ppm. There is no impact

on the operability of plant systems or equipment. Therefore, the change

does not create a malfunction that is different from those previously

evaluated.

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

The proposed increase in the minimum boron concentration in the BAT

provides conservatism in the calculated shutdown margin for Millstone

Unit No. 3. The change does not adversely affect any equipment credited

in the safety analysis. Also, the change does not adversely affect the

probability or consequences of any plant accident, including the

calculated PCT [peak clad temperature] or offsite doses. Therefore,

there is no impact on the margin of safety as specified in the

Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: Phillip F. McKee.

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

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

Minnesota

Date of amendment requests: January 10, 1995.

Description of amendment requests: The proposed amendments would

revise the Prairie Island Event Monitoring Instrumentation Technical

Specifications and associated Bases to conform to Standard Technical

Specifications for post-accident monitoring.

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[s] will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

The primary purpose of post accident monitoring instrumentation is

to display plant variables that provide information to the control room

operators during accident situations. Plant instrumentation was

evaluated for importance for this function when Regulatory Guide 1.97

[''Instrumentation for Light Water Cooled Nuclear Power Plants to

Assess Plant Conditions During and Following an Accident'']

classifications were determined. The Prairie Island Regulatory Guide

1.97 classification of instruments was previously approved by the NRC

on October 18, 1985. This amendment request proposes to base Prairie

Island Technical Specifications on the results of the Regulatory Guide

1.97 evaluation in accordance with the guidance of the industry

standard.

Revising the allowed outage time for these instruments will not

significantly increase the probability or consequences of an accident

since these instruments do not initiate automatic actions, there are

available backup indications and the probability of an event requiring

these instruments to be operable is very low.

Therefore, the probability or consequences of an accident

previously evaluated are not affected by any of the proposed

amendments.

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

new or different kind of accident from any accident previously

analyzed.

The license amendment request proposes to add instruments to the

Technical Specifications which have been previously determined to be

important for post accident monitoring, and to remove instruments from

Technical Specifications which have been previously determined to be

less important for post accident monitoring. This amendment ensures the

control room operators are provided with the instrumentation required

to properly manage an accident situation.

Therefore, based on the above considerations, the possibility of a

new or different kind of accident from any accident previously

evaluated would not be created.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The post accident monitoring functions do not initiate any

automatic actions. The instrumentation to be added to the Event

Monitoring Instrumentation Table was previously recognized through the

Regulatory Guide 1.97 evaluation process as important for post accident

monitoring and would be relied upon if there were an event without this

license amendment. Instrumentation to be removed from Technical

Specifications was previously recognized to be less

[[Page 8754]] important and would not be relied upon very much in an

event. Overall, with the trade-off of adding and deleting

instrumentation, the margin of safety will not be significantly

affected.

The proposed license amendment will increase the allowed outage

time for most of the instruments. Again, these instruments do not

provide automatic actions, they provide indications for monitoring post

accident conditions. All of the instruments have backup or

corroborating indications which could be relied upon if the Technical

Specifications instruments were inoperable. Also, an event requiring

use of these instruments has a very low probability. For these reasons

the proposed changes in allowed outage time will not result in a

significant reduction in the margin of safety.

For these same reasons, the proposed changes in radiation

instrument surveillance requirements will not significantly reduce the

margin of safety.

Overall, a significant reduction in the margin of safety would not

result from this license 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 requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: January 25, 1995.

Description of amendment request: The proposed Technical

Specification change would replace a specific requirement for the

frequency of Type A tests with a general requirement to perform Type A

tests. The proposed amendment would change Surveillance Requirement

4.6.1.2.a. Specifically, the change would require the performance of

Type A tests (overall containment integrated leak rate tests (ILRTs))

at intervals as specified in 10 CFR 50, Appendix J, instead of on a

specific schedule for performance of ILRTs of ``40 plus or minus 10

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. 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 change does not involve a significant increase in the

probability or consequences of an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed change merely replaces a prescriptive

schedule for performing ILRTs with a requirement to conduct the ILRTs

on a schedule consistent with the Commission's regulations. The change

does not alter the methodology, frequency, or acceptance criteria for

ILRTs, does not affect the design basis of the containment, and does

not change the post-accident response of the containment.

B. The change does not create the possibility of a new or different

kind of accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because the change does not affect the manner by which the

facility is operated and does not make any changes to existing plant

structures, systems, or components. The proposed change merely replaces

a prescriptive schedule for performing ILRTs with a requirement to

conduct the ILRTs on a schedule consistent with the Commission's

regulations.

C. The change does not involve a significant reduction in a margin

of safety (10 CFR 50.92(c)(3)) because the proposed change does not

affect the manner by which the facility is operated or involve changes

to equipment or features which affect the operational characteristics

of the facility.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Exeter Public Library, 47

Front Street, Exeter, NH 03833.

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

International Place, Boston, MA 02110-2624.

NRC Project Director: Phillip F. McKee.

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

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: December 7, 1994.

Description of amendments request: The amendments would provide a

permanent voltage-based steam generator tube repair criteria for both

units. This criteria is based on the guidance contained in the NRC

Proposed Generic Communication (Generic Letter 94-XX), ``Voltage-Based

Repair Criteria for the Repair of Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress Corrosion Cracking,'' that was

issued for public comment in the Federal Register (59 FR 41520) on

August 12, 1994. The licensee's submittal also includes responses to

and identifies exceptions taken to the draft Generic Letter. The

significant exceptions are: (1) The requirement to reinspect all tubes

if bobbin probe wear exceeds 15%; (2) the 1 x 10-2 limit on the

calculated conditional burst probability; and (3) the need to pull

additional steam generator tubes to evaluate the current condition of

the steam generator tubes. In addition, the operational leakage

requirement for Unit 2 will be modified to reduce the total allowable

primary-to-secondary leakage for any steam generator from 500 gallons

per day (gpd) to 150 gallons per day.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of Farley units in accordance with the proposed

license amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Testing of model boiler specimens for free standing tubes at room

temperature conditions shows burst pressures as high as approximately

5000 psi for indications of outer diameter stress corrosion cracking

with voltage measurements as high as 26.5 volts. Burst testing

performed on pulled tubes with up to 7.5 volt indications show burst

pressures in excess of 5900 psi at room temperature. As stated earlier,

tube burst criteria are inherently satisfied during normal operating

conditions by the presence of the tube support plate. Furthermore,

correcting for the effects of temperature on material properties and

minimum strength levels (as the burst testing was [[Page 8755]] done at

room temperature), tube burst capability significantly exceeds the R.G.

[Regulatory Guide] 1.121 criterion requiring the maintenance of a

margin of 1.43 times the steam line break pressure differential on tube

burst if through-wall cracks are present without regard to the presence

of the tube support plate. Considering the existing data base, this

criterion is satisfied with bobbin coil indications with signal

amplitudes over twice the 2.0 volt voltage-based repair criteria,

regardless of the indicated depth measurement. This structural limit is

based on a lower 95% confidence level limit of the data. The 2.0 volt

criterion provides an extremely conservative margin of safety to the

structural limit considering expected growth rates of outside diameter

stress corrosion cracking at Farley. Alternate crack morphologies can

correspond to a voltage so that a unique crack length is not defined by

a burst pressure to voltage correlation. However, relative to expected

leakage during normal operating conditions, no field leakage has been

reported from tubes with indications with a voltage level of under 7.7

volts for 3/4 inch tube which correlates to 10 volts for 7/8 inch

tubing (as compared to the 2.0 volt proposed voltage-based tube repair

limit). Thus, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident.

Relative to the expected leakage during accidents (sic) condition

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

leakage and steam release to the environment are Loss of External

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

Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube

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

Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe

Failure is the most limiting for Farley in considering the potential

for off-site doses. The offsite doses analyses for the other events

which model primary-to-secondary leakage and steam releases from the

secondary side to the environment assume that the secondary side

remains intact. The steam generator tubes are not subjected to a

sustained increase in differential pressure, as is the case following a

steam line break event. This increase in differential pressure is

responsible for the postulated increase in leakage and associated

offsite doses following a steam line break event. In addition, the

steam line break event results in a bypass of containment for steam

generator leakage. Upon implementation of the voltage-based repair

criteria, it must be verified that the expected distributions of

cracking indications at the tube support plate intersections are such

that primary-to-secondary leakage would result in site boundary dose

within the current licensing basis. Data indicate that a threshold

voltage of 2.8 volts could result in through-wall cracks long enough to

leak at steam line break conditions. Applications of the proposed

repair criteria requires that the current distribution of a number of

indications versus voltage be obtained during the refueling outages.

The current voltage is then combined with the rate of change in voltage

measurement and a voltage measurement uncertainty to establish an end

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

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

further increased by a factor related to the probability of detection

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

leakage for degraded intersections planned to be left in service

coupled with the reduced specific activity levels allowed result in

radiological consequences outside the current licensing basis, then

additional tubes will be plugged or repaired to reduce steam line break

leakage potential to within the acceptance limit. Thus, the

consequences of the most limiting design basis accident are constrained

to present licensing basis limits.

(2) The proposed license amendment does not create the possibility

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

evaluated.

Implementation of the proposed voltage-based tube support plate

elevation steam generator tube repair criteria does not introduce any

significant changes to the plant design basis. Use of the criteria does

not provide a mechanism that could result in an accident outside of the

region of the tube support plate elevations. Neither a single or

multiple tube rupture event would be expected in steam generator in

which the repair criteria have been applied during all plant

conditions. The bobbin probe signal amplitude repair criteria are

established such that operational leakage or excessive leakage during a

postulate steam line break condition is not anticipated. Southern

Nuclear has previously implemented a maximum leakage limit of 140/150

gpd (Unit 1/Unit 2) per steam generator. The R.G. 1.121 criterion for

establishing operational leakage limits that require plant shutdown are

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

before potential tube rupture. The 140/150 gpd limit provides for

leakage detection and plant shutdown in the event of the occurrence of

an unexpected single crack resulting in leakage that is associated with

the longest permissible crack length. R.G. 1.121 acceptance criteria

for establishing operating leakage limits are based on leak-before-

break considerations such that plant shutdown is initiated if the

leakage associated with the longest permissible crack is exceeded. The

longest permissible crack is the length that provides a factor safety

of 1.43 against bursting at steam line break pressure differential. A

voltage amplitude of approximately 9 volts for typical outside diameter

stress corrosion cracking corresponds to meeting this tube burst

requirement at the 95% prediction interval on the burst correlation.

Alternate crack morphologies can correspond to a voltage so that a

unique crack length is not defined by the burst pressure versus voltage

correlation. Consequently, typical burst pressure versus throughwall

crack length correlations is used below to define the ``longest

permissible crack'' for evaluating operating leakage limits.

The single through-wall crack lengths that results in tube burst at

1.43 times steam line break pressure differential and steam line break

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

leakage for these crack lengths would range from about 0.4 gallons per

minute to 4.5 gallons per minute, respectively, while lower 95%

confidence level leak rates would range from about 0.06 gallons per

minute to 0.6 gallons per minute, respectively.

An operating leak rate of 140/150 gpd per steam generator has been

implemented. This leakage limit provides for detection of 0.4 inch long

cracks at nominal leak rates and 0.6 inch long cracks at the lower 95%

confidence level leak rates. Thus, the 140/150 gpd limit provides for

plant shutdown prior to reaching critical crack lengths for steam line

break conditions at leak rates less than 95% confidence level and for

three times normal operating pressure differential at less than nominal

leak rates.

Considering the above, the implementation of voltage-based plugging

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

accident from any previously evaluated.

(3) The proposed license amendment does not involve a significant

reduction in margin of safety.

The use of the voltage-based tube support plate elevation repair

criteria is demonstrated to maintain steam [[Page 8756]] generator tube

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

describes a method acceptable to the NRC staff for meeting GDCs

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

probability of the consequences of steam generator tube rupture. This

is accomplished by determining the limiting conditions of degradation

of steam generator tubing, as established by inservice inspection, for

which tubes with unacceptable cracking should be removed from service.

Upon implementation of the criteria, even under the worst case

conditions, the occurrence of outside diameter stress corrosion

cracking at the tube support plant elevations is not expected to lead

to a steam generator tube rupture event during normal or faulted plant

conditions. The most limiting effect would be a possible increase in

leakage during a steam line break event. Excessive leakage during a

steam line break event, however, is precluded by verifying that, once

the criteria are applied, the expected end of cycle distribution of

crack indications at the tube support plate elevations would result in

minimal, and acceptable primary to secondary leakage during the event

and, hence, help to demonstrate radiological conditions are less than

an appropriate fraction of the 10 CFR 100 guideline.

The margin to burst for the tubes using the voltage-based repair

criteria is comparable to that currently provided by existing technical

specifications.

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

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

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

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

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

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

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

resulting pressure differential on the deformed tubes may cause some of

the tubes to collapse.

There are two issues associated with steam generator tube collapse.

First, the collapse of steam generator tubing reduces the RCS [reactor

coolant system] flow area through the tubes. The reduction in flow area

increases the resistance to flow of steam from the core during a LOCA

which, in turn, may potentially increase Peak Clad Temperature (PCT).

Second, there is a potential the partial through-wall cracks in tubes

could progress to through-wall cracks during tube deformation or

collapse or that short through-wall indications would leak at

significantly higher leak rates than included in the leak rate

assessments.

Consequently, a detailed leak-before-break analysis was performed

and it was concluded that the leak-before-break methodology (as

permitted by GDC 4) is applicable to the Farley reactor coolant system

primary loops and, thus, the probability of breaks in the primary loop

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

structural design basis of the plant. Excluding breaks in the RCS

primary loops, the LOCA loads from the large branch line breaks were

analyzed at Farley and were found to be of insufficient magnitude to

result in steam generator tube collapse or significant deformation.

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

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

be minimal (much less than 1%) and PCT margin is available to account

for this potential effect. Based on analyses' results, no tubes near

wedge locations are expected to collapse or deform to the degree that

secondary to primary in-leakage would be increased over current

expected levels. For all other steam generator tubes, the possibility

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

not significant. In actuality, the amount of secondary-to-primary

leakage in the event of a LOCA + SSE is expected to be less than that

previously allowed, i.e., 500 gpd per steam generator. Furthermore,

secondary-to-primary in-leakage would be less than primary-to-secondary

leakage for the same pressure differential since the cracks would tend

to tighten under a secondary-to-primary pressure differential. Also,

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

of in-leakage.

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

repair criteria is supplemented by 100% inspection requirements at the

tube support plate elevations having outside diameter stress corrosion

cracking indications, reduced operating leakage limits, eddy current

inspection guidelines to provide consistency in voltage normalization,

and rotating pancake coil inspection requirements for the larger

indications left in service to characterize the principle degradation

mechanism as outside diameter stress corrosion cracking.

As noted previously, implementation of the tube support plate

elevation repair criteria will decrease the number of tubes that must

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

of steam generator tube plugs or tube sleeves would reduce the RCS flow

margin, thus implementation of the voltage-based repair criteria will

maintain the margin of flow that would otherwise be reduced through

increased tube plugging or sleeving.

Considering the above, it is concluded that the proposed change

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

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

bases of the plant Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

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

Alabama 35201

NRC Project Director: William H. Bateman.

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

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: January 9, 1995.

Description of amendments request: The requested changes to the

Technical Specifications (TS) would implement the recommended changes

from Generic Letter 93-05, ``Line Item Technical Specification

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation.'' Specifically, the amendments would implement TS

changes corresponding to the following GL 93-05 line-item improvement

issues: Control Rod Movement Test for Pressurized Water Reactors,

Radiation Monitors, Surveillance of Boron Concentration in the

Accumulator/Safety Injection/Core Flood Tank, Containment Spray System,

Hydrogen Recombiner, and Special Test 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 changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes do not involve any change to the configuration or

method [[Page 8757]] of operation of any plant equipment used to

mitigate the consequences of an accident. The changes to the

surveillance requirements will result in an overall improvement in

plant safety by reducing the likelihood of plant trips and subsequent

challenges to safety systems, decreasing equipment degradation due to

excessive testing, reducing radiation exposure to plant personnel,

increasing the availability of safety related equipment, and

eliminating an unnecessary burden on plant personnel. Therefore, the

proposed changes do 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. The

proposed changes do not involve any change to the configuration or

method of operation of any plant equipment used to mitigate the

consequences of an accident. The relaxation of surveillance tests

curtails the excessive amount of testing that increases wear on the

equipment and reduces the likelihood of plant trips and subsequent

challenges to safety systems. The relaxation also increases the

availability of safety related equipment. Accordingly, no new failure

modes have been defined for any plant system or component important to

safety nor has any new limiting failure been identified as a result of

the proposed changes. 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 changes do not involve a significant reduction in a

margin of safety. The proposed changes eliminate an unnecessary burden

without compromising protection for public health and safety. The

proposed changes were generically analyzed by the NRC as part of a

comprehensive study and presented in NUREG-1366 ``Improvement to

Technical specifications (sic) Surveillance Requirements.'' The NRC

concluded that while some testing at power is essential to verify

equipment and system operability, safety can be improved, equipment

degradation decreased, and unnecessary personnel burden relaxed by

reducing the amount of testing at power. SNC has analyzed plant

operations and made a comparison with the criteria stated in NUREG-1366

for the line-item improvements contained in this request and has found

the NUREG-1366 basis to be consistent with the Farley design and

operation experience. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

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

Alabama 35201.

NRC Project Director: William H. Bateman.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: December 6, 1994.

Description of amendment request: The proposed change to Technical

Specification 3/4.1.3.2 will delete Surveillance Requirement (SR)

4.1.3.2.2, that presently requires, every 31 days, the movement of at

least 2% of its height for each Axial Power Shaping Rod not fully

withdrawn. The proposed amendment would also change the surveillance

intervals for the following Technical Specifications (TS) in accordance

with the guidance of 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:''

1. TS 4.1.3.2 for the Movable Control Assemblies ``Group Height--

Safety and Regulating Rod Groups,'' will relax testing requirements

from at least once every 31 days to every 92 days.

2. TS 4.4.6.2, for ``Operational Leakage,'' relaxes the requirement

to leakage test RCS pressure isolation valves prior to MODE 2 whenever

the plant has been in COLD SHUTDOWN for 72 hours to whenever the plant

has been in COLD SHUTDOWN for 7 days.

3. SR 4.5.2.c.2 for TS 4.5.2, ``ECCS Subsystems--Tavg equal to or

greater than 280 deg. F,'' relaxes the inspection requirements for

ensuring no debris in containment from ``at the completion of each

containment entry'' to ``at least once daily.''

4. TS 4.6.2.1.d, for the ``Containment Spray System,'' relaxes the

SR to perform an air or smoke flow test through the spray header and

nozzles from once per 5 years to once per 10 years.

5. TS 4.10.4.2 for ``Special Test Exceptions Shutdown Margin''

relaxes the SR interval for testing rod insertion capability prior to

reducing shutdown margin from 24 hours to 7 days.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the NRC 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 change does not involve a significant increase in the

probability of an accident previously evaluated nor does it involve a

significant increase in the consequences of an accident previously

evaluated because no change is being made to any accident initiator and

no accident conditions or assumptions used in evaluating the

radiological consequences of an accident are changed. Relaxation of

surveillance requirements is in accordance with GL 93-05, NUREG-1366,

and is compatible with plant operating experience. Deletion of SR

4.1.3.2 is consistent with NUREG-1430, ``Improved Standard Technical

Specifications for B&W Plants.'' No credit is taken in any accident

analysis or mitigation requirements for the Axial Power Shaping Rod

Group.

(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 do not create the possibility of any new or

different kind of accident from any accident previously evaluated

because no new accident initiators or assumptions are introduced by

these proposed changes. Relaxation of SRs as discussed in GL 93-05 was

evaluated as reducing equipment degradation with no increase in safety

consequences consistent with the maintenance of plant specific

reliability of the equipment and systems affected. Deletion of the SR

to move the Axial Power Shaping Rod Group does not affect the

requirement to verify rod position, and there is no credit taken for

movement of these rods to mitigate an accident.

(3) The proposed changes do not result in a significant reduction

in the margin of safety. [[Page 8758]]

The changes do not involve a significant reduction in the margin of

safety, because the proposed changes affect only surveillance

requirements, do not affect the function of the components and systems

involved, and do not decrease the estimated equipment or system

reliability.

Based on the NRC staff analysis, it appears that the three

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

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: University of Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

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

NRC Project Director: Leif J. Norrholm.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: December 6, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 4.0.5, ``Applicability'' and its

associated Bases; TS 3/4.1.2.3, ``Reactivity Control Systems--Makeup

Pump--Shutdown; TS 3/4.1.2.4, ``Reactivity Control Systems--Makeup

Pump--Operating; TS 3/4.1.2.6, Reactivity Control Systems--Boric Acid

Pump--Shutdown; and TS 3/4.1.2.7, ``Reactivity Control System--Boric

Acid Pumps--Operating.'' The proposed change would replace the specific

monthly surveillance requirements associated with the makeup pumps and

boric acid pumps with a surveillance requirement referencing TS 4.0.5,

which references Section XI of the American Society of Mechanical

Engineers Boiler and Pressure Vessel Code for quarterly pump testing

requirements. The proposed change to TS 4.0.5 and its associated Bases

would revise the requirement regarding the NRC's approval of relief

requests to be in accordance with the NRC Staff's recommendation

contained in NUREG-1482, ``Guidelines for Inservice Testing at Nuclear

Power Plants.'' Additionally, TS 4.0.5.a.2 which describes historical

requirements for inservice inspection and testing would be deleted and

TS 4.0.5.a.1 would be renumbered as TS 4.0.5.a.

Basis for proposed no significant hazards consideration

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

performed an 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.

Operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in

accordance with these changes, would not involve a significant increase

in the probability of an accident previously evaluated because no

accident initiators, conditions, or assumptions are affected by the

proposed changes to replace the specific monthly surveillance

requirements for the makeup and boric acid pumps with surveillance

requirements referencing TS 4.0.5 (ASME Boiler and Pressure Vessel Code

Section XI requirements) and to delete wording regarding NRC approval

of relief requests. The changes do not involve a significant increase

in the consequences of an accident previously evaluated, because no

accident conditions or assumptions are affected that would increase the

radiological consequences of a 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 do not result in any new accident initiators

nor do they alter any accident scenarios. The changes do not create the

possibility of a different kind of accident from any accident

previously evaluated, because the surveillance requirements for the

makeup and boric acid pumps only affect the testing of existing

components, systems, and functions, and do not introduce any new

requirements.

(3) The proposed changes do not result in a significant reduction

in the margin of safety.

The proposed changes do not reduce or adversely affect the

capabilities or reliability of any plant structures, systems or

components. Relaxation of the surveillance testing interval for the

boric acid and makeup pumps and modifying the testing requirements is

consistent with previous NRC guidance.

Based on this NRC staff evaluation, it appears that the three

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

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: University of Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

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

NRC Project Director: Leif J. Norrholm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: January 13, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) by relocating Tables 3.3-2,

``Reactor Trip System Instrumentation Response Times,'' and 3.3-5,

``Engineered Safety Features Response Times,'' to FSAR Chapter 16,

Section 16.3. The Bases discussion specific to Table 3.3-5 would also

be relocated to FSAR Section 16.3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed revision does not involve a significant hazards

consideration because operation of Callaway Plant with this change

would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Overall protection system performance will remain within the bounds

of the accident analyses documented in FSAR Chapter 15, WCAP-10961-P,

and WCAP-11883 since no changes to the response times or measurement

interval are proposed.

The RTS and ESFAS will continue to function in a manner consistent

with the above analysis assumptions and the plant design basis. As

such, there will be no degradation in the performance of nor an

increase in the number of challenges to equipment assumed to function

during an accident situation.

These Technical Specification revisions do not involve any hardware

changes nor do they affect the probability of any event initiators.

There will be no change to normal plant operating parameters or

accident mitigation capabilities. Therefore, there will be no increase

in the probability or consequences of any accident occurring due to

these changes.

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

from any previously evaluated. [[Page 8759]]

As discussed above, there are no hardware changes associated with

these Technical Specification revisions nor are there any changes in

the method by which any safety-related plant system performs its safety

function. The normal manner of plant operation is unaffected.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result of

these changes. There will be no adverse effect or challenges imposed on

any safety-related system as a result of these changes. Therefore, the

possibility of a new or different type of accident is not created.

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

No response time changes are proposed in this amendment

application; only the document where these limits are listed will be

changed. There will be no effect on the manner in which safety limits

or limiting safety system settings are determined nor will there be any

effect on those plant systems necessary to assure the accomplishment of

protection functions. There will be no impact on DNBR limits, FQ,

F-delta-H, LOCA PCT, peak local power density, or any other margin of

safety.

Based upon the preceding information, it has been determined that

the proposed changes to the Technical Specifications do not involve a

significant increase in the probability or consequences of an accident

previously evaluated, create the possibility of a new or different kind

of accident from any accident previously evaluated, or involve a

significant reduction in a margin of safety. Therefore, it is concluded

that the proposed changes meet the requirements of 10CFR50.92(C) [sic]

and 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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.

NRC Project Director: Leif J. Norrholm.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: December 8, 1994.

Description of amendment request: The proposed amendment would

change Standby Gas Treatment Power Supply Requirements during refueling

operations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

SGTS [Standby Gas Treatment System] DURING REFUELING OPERATIONS

(Specification 3.7.B.1, 3.7.B.3)

1. The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The Standby Gas Treatment System (SGTS) is not the initiator of any

accident. SGTS may be required to operate for a design basis loss of

coolant accident or for a refueling accident in order to mitigate the

consequences of said accident by providing a filtered exhaust path to

minimize the potential release of radioactive material to the environs.

The proposed amendment does not reduce or change the operational

requirements for the SGTS for an accident. The proposed amendment now

clearly defines the operability requirements during refueling

conditions. The proposed amendment further requires the availability of

a second auxiliary power supply in the event that an Emergency Diesel

Generator (EDG) is out of service during refueling operations, not

currently required. We conclude, therefore, that the proposed amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The SGTS is not an accident initiator, therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed amendment will not involve a significant reduction

in a margin of safety. The proposed amendment requires the availability

of a second auxiliary power supply in the event that an EDG is out of

service during refueling operations, not currently required.

Maintaining availability of a specific reliable auxiliary electrical

power source as an alternative to an EDG in this mode provides

assurance that SGTS can, if required, be operated without placing undue

constraints on EDG availability and represents an enhancement that

increases a margin of safety. We conclude, therefore, that the proposed

amendment does not involve a significant reduction in a margin of

safety.

Based on the above discussion, we have determined that this change

does not constitute a significant hazards consideration as defined in

10CFR50.92(c).

LABORATORY CARBON SAMPLE ANALYSIS (Specification 3.7.B.2.b)

1. The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The Standby Gas Treatment System (SGTS) is not the initiator of any

accident. SGTS may be required to operate for a design basis loss of

coolant accident or for a refueling accident in order to mitigate the

consequences of said accident by providing a filtered exhaust path to

minimize the potential release of radioactive material to the environs.

The proposed amendment does not reduce or change the operational

requirements for the SGTS for an accident. The proposed amendment now

clearly defines the operability requirements during the interval

between sample removal and completion of laboratory analysis.

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

or different kind of accident from any accident previously evaluated.

The SGTS is not an accident initiator, therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed amendment will not involve a significant reduction

in a margin of safety. The proposed change does not reduce the

requirements or acceptance criteria for sampling, testing or analysis.

The proposed change only incorporates into the specification an

existing clarification which addresses the determination of operability

during the time between sample removal and completion of laboratory

analysis. The change provides an explicit time limit consistent with

current regulatory criteria for completion of analyses.

Based on the above discussion, we have determined that this change

does not constitute a significant hazards [[Page 8760]] consideration

as defined in 10CFR50.92(c).

TORUS VENT MODE (Specification 4.7 B.2.c)

1. The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The Standby Gas Treatment System (SGTS) is not the initiator of any

accident. SGTS may be required to operate for a design basis loss of

coolant accident or for a refueling accident in order to mitigate the

consequences of said accident by providing a filtered exhaust path to

minimize the potential release of radioactive material to the environs.

The proposed amendment does not reduce or change the operational

requirements for the SGTS for an accident.

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

or different kind of accident from any accident previously evaluated.

The SGTS is not an accident initiator, therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed amendment will not involve a significant reduction

in a margin of safety. The proposed change will incorporate into the

specification an existing clarification. Use of the SGTS filters during

Torus venting results in an insignificant flow through the filters.

Further, maintaining humidity control prevents any adsorber

degradation. Past sample testing on a six month calendar interval when

720 hours operating time has not accumulated has shown no detectable

impact.

Based on the above discussion, we have determined that this change

does not constitute a significant hazards consideration as defined in

10CFR50.92(c).

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: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301.

Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,

One International Place, Boston, Massachusetts 02110-2624.

NRC Project Director: Walter R. Butler.

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.

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

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

Date of application for amendment: November 19, 1994.

Brief description of amendment: The proposed amendment would revise

Section 3.10.8 and the associated Bases of the Indian Point Nuclear

Generating Unit No. 3 Technical Specifications. Specifically, the

proposed revision would reduce the maximum allowable control rod drop

time from 2.4 to 1.8 seconds. The change would remove, for testing

purposes, the allowance for a seismic event (0.6 seconds), which had

been integral to the 2.4 second safety analysis basis. Since a seismic

event cannot be simulated during the rod drop time test, the more

conservative testing acceptance criteria value of 1.8 seconds is needed

to ensure that the plant is within its design basis. This proposed

revision will support control rod testing which is required during

startup from the current outage.

Date of publication of individual notice in Federal Register:

January 20, 1995 (60 FR 4203).

Expiration date of individual notice: February 21, 1995.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,

and 3, Maricopa County, Arizona

Date of application for amendments: November 30, 1994.

Brief description of amendments: These amendments relocate Table

3.3-2, ``Reactor Protective Instrumentation Response Times,'' and Table

3.3-5, ``Engineered Safety Features Response Times,'' of TS 3/4.3.1 and

3/4.3.2, respectively, to the Palo Verde Updated Final Safety Analysis

Report (UFSAR) in accordance with the guidance provided in Generic

Letter 93-08. In addition, the amendments make administrative changes

to two previous TS amendment requests to maintain consistency with the

deletion of Tables 3.3-2 and 3.3-5. The amendments also delete an

obsolete footnote on page 3/4 3-17 of the Palo Verde Unit 2's TS.

[[Page 8761]]

Date of issuance: February 3, 1995.

Effective date: February 3, 1995.

Amendment Nos.: 88, 75 and 59.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

496) The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 3, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

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: November 16, 1994.

Brief description of amendments: The proposed amendments change the

Technical Specifications to revise the wording for the containment

integrated leakage rate testing in Section 3/4.6.1.2 to make it

consistent with the requirements of the BWR-4 Improved Standard

Technical Specifications (NUREG-1433).

Date of issuance: January 26, 1995.

Effective date: January 26, 1995.

Amendment Nos.: 173 and 204.

Facility Operating License Nos. DPR-71 and DPR-62.

Date of initial notice in Federal Register: December 21, 1994 (59

FR 65810).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 26, 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.

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

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

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

County, Illinois

Date of application for amendments: June 13, 1994, as supplemented

on October 7, 1994.

Brief description of amendments: The amendments revise the

administrative controls in Section 6 of the technical specifications

(TS). The changes include: (1) a change to the submittal frequency of

the Radiological Effluent Release Report from semiannually to annually;

(2) changes to the Shift Technical Advisor (STA) description; (3) a

clarification of the Shift Engineer responsibilities; and (4) several

editorial changes.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995.

Amendment Nos.: 69, 69, 59 and 59.

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 1994 (59 FR

53839).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

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

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

Street, Wilmington, Illinois 60481.

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

Plant, Middlesex County, Connecticut

Date of application for amendment: May 17, 1993 as supplemented

October 12, 1994.

Brief description of amendment: The amendment replaces License

Condition 2.C.4, relating to the implementation and maintenance of the

approved Fire Protection Program, in its entirety with a new License

Condition. In conjunction, with this change, and in accordance with GL

86-10, Technical Specification provisions related to the Fire

Protection Program are being deleted and placed in the Updated Final

Safety Analysis Report.

Date of Issuance: February 1, 1995.

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

30 days.

Amendment No.: 179.

Facility Operating License No. DPR-61. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 7, 1993 (58 FR

36432).

The October 12, 1994, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated February 1, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

application for amendments: August 25, 1994, as supplemented November

16, 1994.

Brief description of amendments: The amendments revise Technical

Specification Table 3.3-4, by revising the ``Trip Setpoint'' and

``Allowable Value'' for the 4 kV bus undervoltage grid degraded voltage

relays and the ``Allowable Value'' for the 4 kV undervoltage loss of

voltage/loss of offsite power relays. This revision was submitted in

response to a concern identified by the licensee in their Self-

Initiated Technical Audit and during the electrical distribution system

functional inspection team findings.

Date of issuance: January 20, 1995.

Effective date: To be implemented within 30 days from the date of

issuance.

Amendment Nos.: 127 and 121.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 12, 1994 (59 FR

51619).

The November 16, 1994, letter provided clarifying information that

did not change the scope of the August 25, 1994, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 20, 1995.

No significant hazards consideration comments received: No.

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

Black Street, Rock Hill, South Carolina 29730.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,

Unit No. 1, St. Lucie County, Florida

Date of application for amendment: July 28, 1994.

Brief description of amendment: This amendment revises Technical

Specifications 3/4.4.13 to incorporate Low Temperature Overpressure

Protection requirements similar to those recommended by the NRC staff

via Generic Letter 90-06. [[Page 8762]]

Date of Issuance: January 27, 1995.

Effective Date: January 27, 1995.

Amendment No.: 132.

Facility Operating License No. DPR-67: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42341).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 27, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,

Georgia

Date of application for amendments: February 3, 1994.

Brief description of amendments: The amendments relocate the

requirements of Technical Specification 3/4.7.10, Area Temperature

Monitoring, to section 16.3 of the VEGP Final Safety Analysis Report

(FSAR). With this relocation to the FSAR, GPC plans to clarify the

basis for areas to be monitored and modify these surveillance

requirements. This change is in accordance with NUREG-1431, ``Standard

Technical Specifications, Westinghouse Plants.''

Date of issuance: January 23, 1995.

Effective date: To be implemented within 30 days from the date of

issuance.

Amendment Nos.: 83 and 61.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 2, 1994 (59

FR 45735).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 23, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830.

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

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

Date of application for amendment: May 13, 1992.

Brief description of amendment: The amendment changes the allowable

primary-to-secondary leakage rate, as specified in License Condition

2.c.(8)2, from 0.1 gallons per minute (gpm) to 0.2 gpm.

Date of Issuance: January 31, 1995.

Effective date: January 31, 1995.

Amendment No.: 193.

Facility Operating License No. DPR-50. Amendment revises a License

Condition.

Date of initial notice in Federal Register: October 14, 1992 (57 FR

47137).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated January 31, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: September 26, 1994.

Brief description of amendment: The amendment revised Technical

Specification 3.5.C.1 and 3.5.C.4 to increase the minimum pressure at

which the high pressure coolant injection system is required to be

operable from 113 psig to 150 psig.

Date of issuance: January 25, 1995.

Effective date: January 25, 1995.

Amendment No.: 166.

Facility Operating License No. DPR-46. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 26, 1994 (59 FR

53841). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 25, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: December 22, 1994.

Brief description of amendment: The amendment revised Technical

Specification 1.0.J, definition of limiting conditions for operation,

consistent with the guidance provided in NRC Generic Letter 87-09,

``Sections 3.0 and 4.0 of the Standard Technical Specifications on the

Applicability of Limiting Conditions for Operation and Surveillance

Requirements.''

Date of issuance: February 3, 1995.

Effective date: February 3, 1995.

Amendment No.: 168.

Facility Operating License No. DPR-46. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 3, 1995 (60 FR

153).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 3, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York.

Date of application for amendment: July 21, 1994.

Brief description of amendment: The amendment revises Technical

Specifications 2.2.2, 3.2.8, 4.2.8, and the associated Bases to reduce

the number of reactor head safety valves required operable from 16

valves to 9 valves. The setpoints of the valve groups are unchanged by

this amendment. The amendment requires testing of the safety valves in

accordance with the approved NMP-1 Inservice Test Program.

Date of issuance: January 25, 1995.

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

30 days.

Amendment No.: 152.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45027).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 25, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

Date of application for amendment: October 28, 1994.

Brief description of amendment: The amendment revises Technical

[[Page 8763]] Specification (TS) 1.7, ``CORE ALTERATION,'' to state

that movement or replacement of incore instrumentation is not

considered to be a CORE ALTERATION and that movement of control rods is

not considered a CORE ALTERATION provided there are no fuel assemblies

in the associated core cell. This amendment includes changes to TS 3/

4.9.3, ``Control Rod Position,'' and associated Bases to be consistent

with the revision to TS 1.7. TS 3/4.9.3 is being revised to require

that all control rods be inserted only during loading of fuel

assemblies into the core rather than during CORE ALTERATIONS. These

changes are consistent with the NRC's, ``Improved Standard Technical

Specifications,'' (NUREG-1434).

This amendment also revises Item 1.i.3) of TS Tables 3.3.2-1 and

4.3.2.1-1 to delete the requirement for Reactor Water Cleanup isolation

due to actuation of the Standby Liquid Control System (SLCS) in

OPERATIONAL CONDITION 5. License Amendment No. 48 issued on September

30, 1993, deleted the requirement for the SLCS to be OPERABLE in

OPERATIONAL CONDITION 5; however, due to an oversight, Item 1.i.3) and

associated notations were not deleted from TS Tables 3.3.2-1 and

4.3.2.1-1 as part of License Amendment No. 48. This amendment corrects

that oversight.

Date of issuance: January 20, 1995.

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

30 days.

Amendment No.: 61.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60382).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 20, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: November 14, 1994.

Brief description of amendment: The amendment revises Technical

Specification 4.5.1.e.2.e) to reduce the leak rate test pressure for

the Automatic Depressurization System (ADS) nitrogen receiving tanks

from 385 psig to 365 psig.

Date of issuance: January 31, 1995.

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

30 days.

Amendment No.: 62.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: December 21, 1994 (59

FR 65817).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 31, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

Date of amendment request: January 14, 1994, as modified by letter

dated October 17, 1994.

Description of amendment request: The amendment revises the

Appendix A 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), and to revise the time within which

the Nuclear Safety Audit Review Committee (NSARC) must issue reports

and minutes.

The amendment also incorporated a number of editorial changes to

delete certain items that are no longer applicable; remove

inconsistencies involving the names of systems, equipment and NSARC

function, composition, and use of alternates; and correct the value for

the reactor coolant system volume. Other editorial changes have

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