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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 12, 1997, through May 22, 1997. The last

biweekly notice was published on May 21, 1997 (62 FR 27792).

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 Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, 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 July 7, 1997, 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

[[Page 30630]]

prehearing conference scheduled in the proceeding, but such an amended

petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, 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 20555-0001,

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.

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

Charlevoix County, Michigan

Date of amendment request: April 22, 1997 (supersedes October 15,

1996, request)

Description of amendment request: The proposed amendment would

revise the Big Rock Point Technical Specifications to correct several

administrative and editorial inconsistencies.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are clarifications within the Technical

Specifications, and do not alter the technical content of the

technical specifications. Plant operation or configuration is not

affected. The postulated doses received by the general public and

plant personnel as a direct result of accidents previously

described, are not affected. Plant operation or configuration is not

affected. Therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes are either clarifications to correct

inconsistencies within the Technical Specifications, or corrections

of typographical errors. The proposed changes do not alter the

facility in any way, 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.

The proposed change[s] [do] not affect any margin of safety as

defined by 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: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: John N. Hannon

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

Charlevoix County, Michigan

Date of amendment request: April 30, 1997

Description of amendment request: The proposed amendment would

alter the company name in the Facility Operating License DPR-6 and

Technical Specifications for the Big Rock Point Plant. Specifically,

the proposed amendment would revise the company name from ``Consumers

Power Company'' to ``Consumers Energy Company.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

[[Page 30631]]

The proposed changes alter the company name in the Facility

Operating License and Technical Specifications to reflect the change

from ``Consumers Power Company'' to ``Consumers Energy Company''.

The company will continue to own all of the same assets, will

continue to serve the same customers, and will continue to honor all

existing obligations and commitments.

Since the proposed changes do not alter the technical content of

any Facility Operating License or Technical Specifications

requirements, they do not alter the design, function, or operation

of any plant structure, system, or component.

Therefore, the changes will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes alter the company name in the Facility

Operating License and Technical Specifications to reflect the change

from ``Consumers Power Company'' to ``Consumers Energy Company''.

The company will continue to own all of the same assets, will

continue to serve the same customers, and will continue to honor all

existing obligations and commitments.

Since the proposed changes do not alter the technical content of

any Facility Operating License or Technical Specifications

requirements, they do not alter the design, function, or operation

of any plant structure, system, or component.

Therefore, the changes will not create the possibility of a new

or different kind of accident from any previously evaluated.

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

Since the proposed changes do not alter the technical content of

any Facility Operating License or Technical Specifications

requirements, they do not alter the design, function, or operation

of any plant structure, system, or component.

Therefore, the changes will not involve a reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: John N. Hannon

Duke Power Company, Docket No. 50-413, Catawba Nuclear Station,

Unit 1, York County, South Carolina

Date of amendment request: May 8, 1997

Description of amendment request: The proposed amendment would add

a phrase to the footnote to Section 3.4.1.2 of the Technical

Specifications that would permit all reactor coolant pumps (RCPs) to be

deenergized for up to 4 hours during Mode 3 on a one-time basis.

Currently, the RCPs are permitted to be deenergized for up to 1 hour

during Mode 3. The proposed change would allow the licensee to perform

a natural circulation test using the new steam generators (installed in

late 1996).

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 activity does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed natural circulation test would be performed in Mode

3 with the reactor subcritical. This transient is bounded by the

transient analyzed in UFSAR [Updated Final Safety Analysis Report]

Section 15.2.6, Loss of Non-Emergency AC Power to the Station

Auxiliaries. For this ANS [American Nuclear Society] Condition II

event, the reactor is assumed to be operating at 102% power, the

turbine driven auxiliary feedwater pump is assumed unavailable and

each steam generator is assumed to have 18% of the steam generator

tubes plugged. By contrast, the planned natural circulation test

would be performed with the reactor subcritical, less than 0.1% of

the tubes plugged in each steam generator, and all support systems

such as auxiliary feedwater, operable for the test. Therefore, the

proposed natural circulation test would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2) The activity does not create the possibility of a new or

different type of accident from any accident previously evaluated.

The proposed change does not involve a physical alteration of

the unit (i.e., no new or different equipment will be installed),

nor will the function of equipment be changed. The change will allow

for a one time performance of a natural circulation test in Mode 3

which will provide useful data on the natural circulation

capabilities of the new Babcock and Wilcox International (BWI) steam

generators that were recently installed at Catawba Unit 1. The test

data will be utilized to validate analysis and simulator models.

Plant operators will also receive valuable experience from

performance of the test. The test will be conducted using written

and approved procedures. An Emergency procedure (EP/1/A/5000/ECA-

0.1) is also available to the Operators for this test. This test is

bounded by the Loss of Non-Emergency AC Power to the Station

Auxiliaries event in Section 15.2.6 of the Catawba UFSAR. For these

reasons, the planned natural circulation test will not create the

possibility of a new or different type of accident from any

previously evaluated.

3) The activity does not involve a significant reduction in the

margin of safety.

Margin of safety is associated with confidence in the ability of

the fission product barriers (the fuel and fuel cladding, the

Reactor Coolant System pressure boundary, and the containment) to

limit the level of radiation doses to the public. As demonstrated by

the bounding UFSAR analysis in Section 15.2.6, none of the fission

product barriers are adversely impacted by the proposed one-time

change. The proposed change does not alter the manner in which

safety limits, limiting safety system setpoints, or limiting

conditions for operation are determined. For these reasons, the

activity 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

proposed amendments involve 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

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

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: February 24, 1997, as supplemented on

April 24, 1997.

Description of amendment request: The licensee proposed changes to

Technical Specification Section 6.9.1.7, Core Operating Limits Report,

to reflect use of the Westinghouse Best Estimate Large Break Loss-of-

Coolant Accident (LOCA) methodology for large break LOCA analysis,

including supporting documents.

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:

Question 1 Does the proposed license amendment involve a

significant increase in the probability or consequences of an

accident previously evaluated?

The plant conditions assumed in the analysis are bounded by the

design conditions for all equipment in the plant. Therefore, there

will be no increase in the probability of a Loss of Coolant Accident

[[Page 30632]]

(LOCA). The consequences of a LOCA are not being increased. That is,

it is shown that the emergency core cooling system is designed so

that its calculated cooling performance conforms to the criteria

contained in 10 CFR 50.46 paragraph (b). No other accident is

potentially affected by this change. Therefore, neither the BiWeekly

probability nor the consequences of an accident previously evaluated

is increased due to the proposed change.

Question 2 Does the proposed license amendment create the

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

previously evaluated?

No new modes of plant operation are being introduced. The

parameters assumed in the analysis are within the design limits of

existing plant equipment. All plant systems will perform as designed

in response to a potential accident. Therefore, the proposed license

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

of accident from any accident previously evaluated.

Question 3 Does the proposed amendment involve a significant

reduction in the margin of safety?

The analysis in support of the proposed license amendment

realistically models the expected response of the Turkey Point Units

3 & 4 nuclear core during a postulated LOCA. Uncertainties have been

accounted for as required by 10 CFR 50.46. A sufficient number of

loss of coolant accidents with different break sizes, different

break locations and other variations in properties have been

calculated to provide assurance that the most severe postulated loss

of coolant accidents were analyzed. It has been shown by the

analysis that there is a high level of probability the criteria

contained in 10 CFR 50.46 paragraph (b) would not be exceeded.

Therefore, the proposed amendment 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: February 17, 1997 as revised May 1,

1997.

Description of amendment request: The proposed amendment would

change the Crystal River Unit 3 (CR-3) Technical Specifications (TS) to

implement 10 CFR Part 50, Appendix J, ``Primary Reactor Containment

Leakage Testing for Water-Cooled Reactors,'' Option B. This option

allows to change from prescriptive testing requirements to performance-

based testing requirements based on the leakage rate testing history of

the containment and components. The proposed TS changes include

revision to TS 3.6.1, 3.6.3, and addition of ``Containment Leakage Rate

Testing Program'' to TS 5.0. The licensee did not propose any

deviations from methods approved by the Commission and endorsed in the

applicable regulatory guide. This notice supersedes the previous notice

dated February 28, 1997 (62 FR 9214)

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 TS amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes to the TS are to implement Option B of 10

CFR 50, Appendix J, at CR-3. The proposed changes will result in

increased intervals between containment leakage tests based on the

leakage rate testing history. The proposed changes do not involve a

change to the plant design or operation and does not change the

testing methodology.

NUREG-1493, ``Performance-Based Containment Leak-Test Program,''

provides the technical basis of 10 CFR 50, Appendix J, Option B.

NUREG-1493 contains a detailed evaluation of the expected leakage

from containment and the associated consequences. The increased risk

due to increasing the intervals between containment leakage tests

was also evaluated. The NUREG used a statistical approach to

determine that the increase in the expected dose to the public due

to decreasing the testing frequency is extremely low. NUREG-1493

also concluded that a small increase is justifiable in comparison to

the benefits from decreasing the testing frequency. The primary

benefit is in the reduction in occupational radiation exposure.

Criterion 2

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

accident from any accident previously evaluated.

The TS amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed TS amendment incorporates the performance-based

testing approach authorized by 10 CFR 50 Appendix, J, Option B.

Decreasing the testing frequency allowed by this change does not

involve a change to plant design or operation. Safety related

equipment and safety functions are not altered as a result of this

change. Decreasing the testing frequency does not affect testing

methodology. As a result, the proposed change does not affect any of

the parameters or conditions that could contribute to the initiation

of any accidents.

Criterion 3

Does not involve a significant reduction in the margin of

safety.

This TS amendment does not involve a significant reduction in

the margin of safety.

The proposed TS amendment does not change the methodology of the

containment leakage rate testing program or program acceptance

criteria. The proposed TS change does affect the frequency of

containment leakage rate testing. With an increased interval between

tests, a small possibility exists that an increase in leakage could

go undetected for a longer period of time. Based on the operational

experience at CR-3, it has been demonstrated that the leak-tightness

of the containment building has consistently been significantly

below the allowable leakage limit. Adequate controls are in place to

ensure that required maintenance and modifications are performed.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Attorney for licensee: R. Alexander Glenn, Corporate Counsel,

Florida Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg,

Florida 33733-4042

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: March 27, 1997, as supplemented April 3,

and May 1, 1997.

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) for the Crystal River Nuclear

Plant Unit 3 (CR3) relating to the Once Through Steam Generator's

(OTSG's) tube inspection acceptance criteria. Specifically, the

licensee proposed to:

(1) revise TS 3.4.12 (d) to specify 150 gallons per day limit on

primary-to-secondary leakage through either OTSG;

(2) add TS 5.6.2.10.2 e. to define inspection requirements and

disposition criteria for applicable tubes in the ``B'' OTSG first span;

[[Page 30633]]

(3) revise TS 5.6.2.10.4.a.7 to define ``pit-like Intergranular

attack indications

(4) revise TS 5.6.2.10 and 5.7.2 to delete requirements that were

specific to the interim tube plugging criteria applicable until Refuel

11.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

FPC Response:

No. The CR-3 components addressed by this proposed change are

the Once Through Steam Generators (OTSGs), identified by plant

tagging procedures as RCSG-1A and RCSG-1B. The OTSGs are straight

tube, straight shell heat exchangers which allow for heat removal

and the subsequent production of steam as a result of heat transfer

from the primary side reactor coolant to the secondary side

feedwater. Proposed changes are; retaining reduced primary-to-

secondary leak rates approved previously for one cycle only,

returning inspection result reporting requirements to those

previously implemented, and establishing new inspection requirements

for the ``B'' OTSG. Sunset clauses are being removed from pages

containing requirements effective for one refueling outage and

subsequent operating cycle only.

Based on review of Chapter 14 of the CR-3 Final Safety Analysis

Report (FSAR), FPC performed analyses to assess the consequences of

a steam generator tube rupture event, including the complete

severing of a steam generator tube. The analyses concluded that CR-3

was sufficiently designed to ensure that, in the event of a steam

generator tube rupture, the radiological doses would not exceed the

allowable limits prescribed by 10 CFR 100, and would not result in

additional tube failures and further degradation of the reactor

coolant pressure boundary.

Retaining the present primary-to-secondary leakage limit (LCO

3.4.12, RCS Operational Leakage) that was previously approved for

the current operating cycle will continue to provide assurance that

should a significant leak occur, it would be detected and the plant

will be shut down in a timely manner to reduce the likelihood of a

potential tube rupture. This value of primary-to-secondary leakage

applicable to both OTSGs is conservative relative to existing safety

analyses and would result in lower doses than currently calculated

and found acceptable. Removing reporting requirements specific to

use of alternate flaw sizing criteria approved for Refueling Outage

10 only, and returning to previous reporting requirements applicable

to both OTSGs, has no effect on operating plant safety. These

requirements are administrative only and do not affect steam

generator inspection or disposition of inspection results.

The proposed change to the ``B'' OTSG inspection criteria

establishes that future inspections will include 100% inspection of

the first span of specific tubes which are known to have indications

of degradation. The degradation of these tubes is attributed to a

common non-random mechanism.

The results of inspections of these tubes will be dispositioned

using the same criteria as all other OTSG tubes for determination of

the need for plugging or sleeving. Therefore, the proposed change

will not increase the probability or consequence of an accident

previously evaluated as all tubes degraded beyond acceptable limits

will be subject to consistent corrective actions.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different kind of

accident from any accident previously evaluated?

FPC Response:

No. The purpose of OTSG tube inspection is to identify tubes

that may have a higher potential for failure due to degradation that

results in a reduced ability to withstand operating conditions.

Neither the type of inspection of OTSG tubes nor the process for

performing inspections will be changed by this amendment. Consistent

criteria will be applied to disposition inspection results and

consistent corrective actions will be taken for tubes that exceed

this criteria. Retaining the lower leakage limit is conservative

relative to existing analyses. Changes to revise requirements for

reporting inspection results, and remove ``sunset'' clauses

addressing the applicability of License Amendment 154 until

Refueling Outage 11 only, do not alter the design or operation of

the OTSGs. Therefore, no new or different kind of accident will be

created as a result of these changes.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in margin of safety?

FPC Response:

No. The analyses that have been performed on the effects of OTSG

tube failures, as reported in the CR-3 FSAR, have demonstrated that

internal and offsite consequences are within allowable limits. This

change will not alter the acceptance criteria for inspection

results. Since this change will assure that a group of tubes with

existing first span pit-like inter-granular attack indications are

inspected each inspection period, the likelihood of detecting active

degradation, as well as the probability of repairing degraded tubes

prior to the degradation resulting in a through-wall opening or tube

rupture, is increased. Retaining the currently accepted primary-to-

secondary leakage limit continues to provide assurance that should a

significant leak occur, it would be detected and the plant will be

shut down in a timely manner to reduce the likelihood of a potential

tube rupture, thereby maintaining or improving the existing margin

of safety. Changes to revise requirements for reporting inspection

results, and remove ``sunset'' clauses addressing the applicability

of License Amendment 154 until Refueling Outage 11 only, do not

alter the design or operation of the OTSGs. Therefore, these changes

will not involve a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Attorney for licensee: R. Alexander Glenn, Corporate Counsel,

Florida Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg,

Florida 33733-4042

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: May 8, 1997

Description of amendment request: The proposed amendment

incorporates additional analytical methods, GPU Nuclear Topical

Reports, TR-078, TR-087, TR-091, and TR-092P, previously approved by

the NRC, to Technical Specifications (TS) Section 6.9.5.2. These

Topical Reports will be utilized by GPU Nuclear to perform core reload

design analysis for the Three Mile Island, Unit 1 (TMI-1) Facility. TS

6.9.5.2 is also being editorially revised to relocate the existing note

that the current revision level shall be specified in the Core

Operating Limits Report (COLR) such that it applies to the additional

Topical Reports, as well as BAW-10179 P-A.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

GPU Nuclear has determined that this Technical Specification

Change Request poses no significant hazards as defined by 10 CFR

50.92.

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed change to reference the

analytical methodologies specified in GPU Nuclear Topical Reports

TR-078, TR-087, TR-091,and TR-092 use[d] in TMI-1 core reload design

analysis is considered administrative since these Topical Reports

[[Page 30634]]

have been reviewed and approved by the NRC for use at TMI-1.

Therefore, the proposed change does not involve a significant

increase in the probability of occurrence or the consequences of an

accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. The proposed change

to reference NRC-approved GPU Nuclear Topical Reports TR-078, TR-

087, TR-091, and TR-092P will continue to ensure that approved

methods and criteria are used to establish core operating limits.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed change to reference NRC-approved GPU Nuclear

Topical Reports TR-078, TR-087, TR-091, and TR-092P maintains

existing margins of safety since approved methods and criteria are

still used to establish core operating limits.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: Patrick D. Milano, Acting

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

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

Connecticut

Date of amendment request: May 8, 1997

Description of amendment request: The proposed amendment would

modify the minimum accuracy stated in Technical Specification (TS)

Table 3.3-8, Meteorological Monitoring Instrumentation,''

for the instruments used to measure wind speed and air temperature -

delta T. TS Bases Section 3/4.3.3.4 would also be modified to reflect

the proposed changes to TS Table 3.3-8.

Regulatory Guide 1.23 (Safety Guide 23), ``Onsite Meteorological

Programs,'' dated March 17, 1972, provides recommended instrument

accuracies for meteorological instrumentation. The proposed minimum

instrument accuracies for the air temperature - delta T and the wind

speed (only when the wind speed is greater than 5 miles per hour) do

not meet the recommended accuracies of Regulatory Guide 1.23. However,

margin is included to account for uncertainties.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes modify the accuracy requirements for the

instruments which are used to measure wind speed and air temperature

- delta T. The data obtained from the meteorological instrumentation

would be used to: a) estimate the public dose following routine or

accidental releases of airborne radioactivity, b) make decisions

regarding actions to protect the public in the event of an accident

involving a release of airborne radioactivity, and c) establish

radiological dispersion parameters to determine radiological doses

in design basis accident calculations.

The proposed minimum instrument accuracy requirements are more

than sufficient to meet the purposes denoted above. The

meteorological parameters measurement uncertainties insignificantly

affect the results when compared to the accuracies of the source

term estimates, meteorological dispersion models, dose models, and

meteorological forecasting. Therefore, there is no impact on the

consequences (offsite doses) associated with previously evaluated

accidents.

The proposed changes do not alter the way any structure, system,

or component functions, do not alter the manner in which the plant

is operated, and do not have any impact on the protective boundaries

and safety limits for the protective boundaries. Therefore, the

proposed changes do not impact the probability of any previously

evaluated accidents.

Thus, the license amendment request does not impact the

probability of an accident previously evaluated nor does it involve

a significant increase in the consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes modify the accuracy requirements for the

instruments which are used to measure wind speed and air temperature

- delta T. The data provided by these instruments assist in

responding to a design basis accident which may involve a release of

airborne radioactivity. The instruments are used for post accident

monitoring and serve a passive role; they cannot initiate or

mitigate any accident.

The proposed changes do not alter the way any structure, system,

or component functions and do not alter the manner in which the

plant is operated. They do not introduce any new failure modes.

Thus, the license amendment request does not create the

possibility of a new or different kind of accident from any

previously analyzed.

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

As discussed above, the proposed changes modify the accuracy

requirements for the instruments which are used to measure wind

speed and air temperature - delta T which could impact the

radiological dispersion coefficient used to determine radiological

doses in design basis accident calculations. However, the

differences in the instrument accuracies and the Regulatory Guide

1.23 requirements have been determined not to significantly affect

the dispersion coefficients. Thus, there is no significant impact on

offsite doses associated with previously analyzed accidents.

Therefore, there is no significant reduction in the margin of safety

for the design basis accident analysis.

Thus, the license amendment request does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: April 14, 1997

Description of amendment request: Technical Specification 3.4.9.3.a

requires two relief valves be operable to protect the reactor coolant

system from overpressurization when any reactor coolant system cold leg

is less than 350F. The proposed amendment revises the setpoint of the

residual heat removal suction relief valves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

[[Page 30635]]

consideration, which is presented below:

NNECO has reviewed the proposed change in accordance with 10CFR

50.92 and has concluded that the change does not involve a

significant hazards consideration (SHC). The bases for this

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

satisfied. The proposed change does not involve a SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to Technical Specification 3.4.9.3 to

decrease the setpoint of the residual heat removal suction relief

valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus]

3% ([greater than or equal to] 426.8 psig and [less than or equal

to] 453.2 psig) is consistent with the design capabilities and

system requirements of the relief valves and the relief valves are

not credited in previously evaluated accidents.

Therefore, the proposed change does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed change to Technical Specification 3.4.9.3 to

decrease the setpoint of the residual heat removal suction relief

valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus]

3% ([greater than or equal to] 426.8 psig and [less than or equal

to] 453.2 psig) does not change the operation of the residual heat

removal system, reactor coolant system or any system component

during normal or accident evaluations. The proposed change to the

setpoint of the residual heat removal suction relief valves from 450

psig [plus or minus] 3% to 440 psig [plus or minus] 3% ([greater

than or equal to] 426.8 psig and [less than or equal to] 453.2 psig)

also ensures protection of the reactor coolant system against cold

overpressurization transients in accordance with the requirements of

10CFR50, Appendix G.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

The proposed change to Technical Specification 3.4.9.3 to

decrease the setpoint of the residual heat removal suction relief

valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus]

3% ([greater than or equal to] 426.8 psig and [less than or equal

to] 453.2 psig) provides an acceptable allowance between the maximum

relief valve setpoint ([less than or equal to] 453.2 psig) and

10CFR50, Appendix G requirements. The proposed change to the

setpoint provides sufficient allowance between the minimum relief

valve setpoint ([greater than or equal to] 426.8 psig) and reactor

coolant system pressure when residual heat removal system is

unisolated from the reactor coolant system to minimize the

probability of an inadvertent residual heat removal system relief

valve opening.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed change does not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: April 28, 1997

Description of amendment request: Technical Specification

Surveillances 4.1.2.3.1, 4.1.2.4.1, 4.5.2.f, and 4.5.2.h require the

charging and safety injection pumps to be tested on a periodic basis

and after modifications that alter subsystem flow characteristics. The

proposed amendment would increase the required differential pressure at

recirculation flow for the safety injection and centrifugal charging

pumps; decrease the required individual safety injection and

centrifugal charging pump injection line flow rate; increase the

allowed individual safety injection pump total flow rate; and make

editorial changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed changes in accordance with

10CFR50.92 and has concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed changes do not involve [an] SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to Technical Specification Surveillances

4.1.2.3.1, 4.1.2.4.1, and 4.5.2.f to increase the required discharge

pressure for the centrifugal charging pumps on recirculation flow

during surveillance testing from [greater than or equal to] 2411

psid to [greater than or equal to] 5676 ft (2464 psid) are

consistent with centrifugal charging pump design requirements. The

change in the referenced units from differential pressure measured

in psid to total head measured in feet for the centrifugal charging

pumps and safety injection pumps during surveillance testing is an

administrative change.

The proposed changes to Technical Specification Surveillance

4.5.2.f to increase the required discharge pressure for the safety

injection pumps on recirculation flow during surveillance testing

from [greater than or equal to] 1348 psid to [greater than or equal

to] 3240 ft (1406 psid) are consistent with safety injection pump

design requirements.

The proposed changes to Surveillance 4.5.2.h: to decrease the

required individual centrifugal charging pump injection line flow

rate sum from [greater than or equal to] 339 gpm to [greater than or

equal to] 310.5 gpm, decrease the required individual safety

injection pump injection line flow rate sum from [greater than or

equal to] 442.5 gpm to [greater than or equal to] 423.4 gpm,

increase the required individual safety injection Pump A total flow

rate from [less than or equal to] 670 gpm to [less than or equal to]

675 gpm, and increase the required individual safety injection Pump

B total flow rate from [less than or equal to]

650 gpm to [less than or equal to] 675 gpm are consistent with

centrifugal charging pump and safety injection pump design

requirements.

The proposed changes are consistent with equipment design

requirements and performing surveillance testing does not involve a

significant increase in the probability of an accident previously

evaluated.

The proposed changes to the surveillance testing of the

centrifugal charging pumps and safety injection pumps provide the

necessary assurance that the pumps will function consistent with the

flows used in the accident analyses and does not involve a

significant increase in the consequence of an accident previously

evaluated.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes to the surveillance testing of the

centrifugal charging pumps and safety injection pumps do not change

the operation of the centrifugal charging or safety injection

systems or any of its components during normal or accident

evaluations. The increase in the allowed maximum safety injection

pump flow does not impact the cold overpressure accident analysis.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

[[Page 30636]]

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

The proposed changes to Technical Specification Surveillances

4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f to increase the required discharge

pressure for the centrifugal charging pumps on recirculation flow

during surveillance testing from [greater than or equal to] 2411

psid to [greater than or equal to] 5676 ft (2464 psid) provides an

acceptable margin between the required surveillance and design pump

performance to provide assurance that the pumps will operate

consistent with the assumptions of the accident analysis.

The proposed changes to Technical Specification Surveillance

4.5.2.f to increase the required discharge pressure for the safety

injection pumps on recirculation flow during surveillance testing

from [greater than or equal to] 1348 psid to [greater than or equal

to] 3240 ft (1406 psid) provides an acceptable margin between the

required surveillance and design pump performance to provide

assurance that the safety injection pumps will operate consistent

with the assumptions of the accident analysis.

The proposed changes to Surveillance 4.5.2.h to decrease the

required individual centrifugal charging pump injection line flow

rate sum from [greater than or equal to] 339 gpm to [greater than or

equal to] 310.5 gpm, decrease the required individual safety

injection pump injection line flow rate sum from [greater than or

equal to] 442.5 gpm to [greater than or equal to] 423.4 gpm,

increase the required individual safety injection Pump A total flow

rate from [less than or equal to] 670 gpm to [less than or equal to]

675 gpm and increase the required individual safety injection Pump B

total flow rate from [less than or equal to] 650 gpm to [less than

or equal to] 675 gpm are consistent with the assumptions of the

accident analysis. The maximum allowed safety injection flow is

consistent with the vendor recommendation for maximum continuous

runout flow. Also, the safety injection

pumps are disabled during specific normal operating modes,

consistent with the assumptions of the accident analysis, to ensure

that they can not be an injection source when the cold overpressure

system is required to be operable and thus the increase in maximum

safety injection pump flow does not affect the cold overpressure

accident analysis.

The change in the referenced units in Technical Specification

Surveillances 4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f from differential

pressure measured in psid to total head measured in feet for the

centrifugal charging pumps and safety injection pumps during

surveillance testing is an administrative change.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed changes do not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: April 28, 1997

Description of amendment request: Technical Specification 3.7.6

requires that flood protection be provided for the service water pump

cubicles and components when the water level exceeds a specific value.

The proposed amendment (1) adds the closing of the service water pump

cubicle sump drain valves, (2) revises the wording of the action

statement to be consistent with the limiting condition for operation,

and (3) revises the associated Bases section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed change in accordance with

10CFR50.92 and has concluded that the change does not involve a

significant hazards consideration (SHC). The bases for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed change does not involve [an] SHC because the

change would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to Technical Specification 3.7.6 identify

additional manual actions to be performed to provide external lood

protection for the service water pump cubicles in the event of high

water level (13 ft MSL) [mean sea level]. The cubicle sump drain

valves which are to be closed are part of a modification which

installed a drain line from the sump of each cubicle to the intake

bay in order to provide a passive means of removing internal leakage

from the cubicle. The cubicle sump drain valves are normally

maintained in the open position.

The drain valves meet the intent of RG [Regulatory Guide] 1.59

for ``hardened protection'' and RG 1.102 for ``incorporated

barriers'' in a manner similar to that of the cubicle watertight

doors. RG 1.59 states that hardened protection ``must be passive and

in place, as it is to be used for flood protection, during normal

plant operation''. RG 1.102 states that ``the plant should be

designed and operated to keep doors necessary for flood protection

closed during normal operation''. The Response to FSAR [Final Safety

Analysis Report] Question No. 240.9 established the acceptability of

the practice of maintaining one service water pump cubicle

watertight door open and the other door closed during normal

operations.

The proposed change in the action statement to initiate action

when water level is exceeding 13 feet MSL rather than at 13 feet MSL

is a clarification only which provides consistency between the

limiting condition for operation and the action statements.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes to Technical Specification 3.7.6 identify

additional, simple to perform manual actions to provide external

flood protection for the service water pump cubicles.

The proposed change in the action statement to initiate action

when water level is exceeding 13 feet MSL rather than at 13 feet MSL

and the proposed changes to the bases are considered clarifications.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

The proposed changes to Technical Specification 3.7.6 identify

additional, simple to perform manual actions to provide external

flood protection for the service water pump cubicles in the

event of high water level (13 ft MSL). The plant modification which

made these additional actions necessary was made to provide for

improved internal flood protection.

The proposed change in the action statement to initiate action

when water level is exceeding 13 feet MSL rather than at 13 feet MSL

and the proposed changes to the bases are considered clarifications.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed change does not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

[[Page 30637]]

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: May 1, 1997

Description of amendment request: Technical Specifications 3/

4.8.2.2 and 3/4.8.3.2 specify which electrical power systems are

required to be operable in Modes 5 and 6. The proposed amendment would

clarify the requirements by identifying the specific equipment required

and their alignments in Modes 5 and 6.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed changes in accordance with 10CFR

50.92 and has concluded that the change does not involve a

significant hazards consideration (SHC). The bases for this

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

satisfied. The proposed changes do not involve [an] SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed change to Technical Specification 3/4.8.2.2 to

replace the wording ``As a minimum, one 125 volt battery bank and

its associated full capacity charger'' to ``As a minimum, one Train

(A or B) of batteries and their associated full capacity

chargers'' will increase the required battery banks operable from

one to two.[]

This change is being proposed to resolve an inconsistency with

Technical Specification 3/4.8.3.2 which currently requires two

battery banks energized in modes 5 and 6.

The proposed change to...Technical Specifications 3/4.8.2.2 and

3/4.8.3.2 to identify the specific equipment required and its

alignment during modes 5 and 6 is being proposed to reduce the

vagueness in the present Technical Specifications. This proposed

change will specify the equipment required operable for the

electrical distribution systems during modes 5 and 6.

These proposed changes are considered administrative and do not

alter the manner in which any system or component is operated or

expected to respond during an accident. Therefore, the proposed

changes do not involve a significant increase in the probability or

consequence of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes to Technical Specification 3/4.8.2.2 to

increase the required battery banks operable from one to two and to

reword Technical Specifications 3/4.8.2.2 and 3/4.8.3.2 to identify

the specific equipment required operable during modes 5 and 6 do not

alter the manner in which any system or component is operated or

expected to respond during normal or accident conditions.

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.

The proposed changes to Technical Specification 3/4.8.2.2 to

increase the required battery banks operable from one to two is

being proposed to resolve an inconsistency with Technical

Specification 3/4.8.3.2 which currently requires two battery banks

energized in modes 5 and 6. This is considered an administrative

change.

The proposed changes to...Technical Specifications 3/4.8.2.2 and

3/4.8.3.2 are being proposed to reduce the vagueness in the present

technical specifications by identifying the specific equipment

required operable during modes 5 and 6. The change will provide a

greater level of assurance that the electrical distribution systems

will be correctly aligned and surveilled. This is also considered an

administrative change.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed changes to not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: May 5, 1997

Description of amendment request: Technical Specification

Surveillance 4.8.4.1 requires periodic testing of lower voltage circuit

breakers for all containment penetration conductor overcurrent

protective devices. The proposed amendment would modify the

requirements for determining the operability of lower voltage circuit

breakers by using the manufacturer's curve of current versus time to

test delay trip elements, clarify the use of two pole in series

testing, and expand the Bases description of the testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The bases for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed revision does not involve [an] SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed change to Technical Specification Surveillance

4.8.4.1 to modify the requirements for determining the operability

of lower voltage circuit breakers by using the manufacture's curve

of current versus time to test long time and short-time delay trip

elements will not change the requirement that periodic testing be

performed to determine breaker operability. The circuit breaker

testing is consistent with the design of the components and

performing surveillance testing does not involve a significant

increase in the probability of an accident previously evaluated. The

proposed change will provide assurance that the breakers will

perform consistent with accident analyses and does not involve a

significant increase in the consequence of an accident previously

evaluated.

The proposed change to the surveillance to modify the wording

associated with the use of two pole in series testing to determine

Molded Case Circuit Breaker (MCCB) operability following the failure

of [an] MCCB to pass a single pole test was previously approved in

License Amendment No. 13. The modified wording clarifies the testing

by specifically stating in the surveillance that the two pole in

series test determines MCCB operability. This is considered an

administrative change.

The proposed change to expand the description of the long-time

and short-time delay trip elements testing in the Bases Section is

also considered an administrative change.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

[[Page 30638]]

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

from any accident previously evaluated.

The proposed change to use a curve of current versus time

instead of the description in Technical Specification Surveillance

4.8.4.1 of the [] long-time and short-time delay trip element

testing does not alter the design, operation, or maintenance of the

lower voltage circuit breakers.

The proposed change to the surveillance to modify the wording

associated with the use of two pole in series testing to determine

MCCB operability and the expanded description of the long-time and

short-time delay elements testing in the Bases Section are

considered administrative changes.

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.

The current wording of Technical Specification Surveillance

4.8.4.1 requires testing of long-time delay trip elements with a

current value of exactly 300% of the pickup setting and short-time

delay trip elements with a current value of exactly 150% of the

pickup setting. The testing [cannot] be performed at exact values.

Circuit breaker manufactures develop a curve of current versus time

for each breaker type that specifies the allowable time to trip for

a specified current. Using the curve for a given breaker type, the

operability of a circuit breaker can be verified by inserting a

given current and verifying that the breaker trips within the

allowable time delay band width for that current. Testing by the

industry is typically performed at approximately 300% of the pickup

setting for long-time delay trip elements and approximately 150% of

the pickup setting for short-time delay trip elements. The proposed

change to the surveillance to modify the requirements for

determining the operability of circuit breakers by using the

manufacturer's curve of current versus time to test delay trip

elements will continue to provide assurance that lower voltage

circuit breakers for all containment penetration conductor

overcurrent protective devices will operate consistent with the

assumptions of the accident analysis.

The proposed change to the surveillance to modify the wording

associated with the use of two pole in series testing to determine

MCCB operability and the expanded description of the long-time and

short-time delay trip elements testing in the Bases Section are

considered administrative changes.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed changes do not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: May 5, 1997

Description of amendment request: Technical Specification

Surveillance 4.5.2.b.1 requires that the emergency core cooling system

(ECCS) piping be verified full of water at least once per 31 days. The

proposed amendment would revise the surveillance to exempt the

operating charging pump(s) and associated piping from the requirement

to be verified full of water and move the description of the

verification method from the surveillance to the Bases section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The bases for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed revision does not involve [an] SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed change to Technical Specification Surveillance

4.5.2.b.1 to exempt the operating centrifugal charging pump(s) and

associated piping from the requirement to be vented will not effect

the requirement the ECCS piping be full of water. An operating

centrifugal charging pump and the associated piping is self venting

and cannot develop voids and pockets of entrained gases. This change

is consistent with the design of the charging system and ensuring

that ECCS piping is full of water does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

The proposed change Technical Specification Surveillance

4.5.2.b.1 to move and expand the description of the venting method

from the surveillance to the Bases Section are considered

administrative changes.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed change to exempt the operating centrifugal charging

pump(s) and associated piping from the requirement to be

periodically vented by crediting its self venting capabilities does

not change the operation of the charging system or any of its

components during normal or accident evaluations.

The proposed changes to move and expand the description of the

venting method from the surveillance to the Bases Section are

considered administrative changes.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

The proposed change to Technical Specification Surveillance

4.5.2.b.1 to exempt the operating centrifugal charging pump(s) and

associated piping from the requirement to be manually vented by

crediting its self venting capabilities, is consistent with the

design of the charging system. This proposed change continues to

ensure that ECCS piping is full of water and thus, does not involve

a significant reduction in a margin of safety.

The proposed change to Technical Specification Surveillance

4.5.2.b.1 to move the description of the venting method from

thesurveillance to the Bases Section is considered an administrative

change. Currently the surveillance identifies that ECCS piping is to

be verified full of water by venting ECCS pump casings and

accessible discharge piping high points except for the RSS

[recirculation spray system] pump, RSS heat exchanger and associated

RSS piping that are not maintained filled with water during plant

operation. The venting description will be expanded when moved to

the bases to include an exclusion for the above described operating

centrifugal charging pump(s) and associated piping and the venting

method used for nonoperating centrifugal charging pumps. The

centrifugal charging pumps have top mounted suction and discharge

nozzles and do not have casing vents. The pump manufacturer has

indicated that venting the pump suction pipe will assure that the

pump is full of water. This venting of the nonoperating centrifugal

charging pumps is accomplished by using a pump suction line test

connection.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed change does not involve an SHC.

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

[[Page 30639]]

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: April 17, 1997

Description of amendment request: This license amendment request

revises Technical Specification (TS) 2.12, ``Control Room System,'' to

delete the Limiting Conditions of Operation (LCO) and associated

surveillance for the control room temperature and replace it with an

LCO and surveillance on the control room air conditioning (A/C) system.

The remainder of TS 2.12 is being rewritten consistent with the

requirements of the Combustion Engineering Standard TS (NUREG-1432,

Rev. 1). In reviewing requirements for refueling and shutdown

operations, additional TS improvement were identified. Therefore, the

definition section, TS 2.1 ``Reactor Coolant System,'' 2.6

``Containment System,'' 2.8 ``Refueling Operations,'' and associated

surveillance requirements are proposed for revision to incorporate the

design basis requirements for refueling operations and to correspond to

NUREG-1432.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will incorporate new requirements for the

control room air conditioning system, control room filtration

system, and refueling operations. In addition, the proposed change

will ensure that the Limiting Condition for Operations and

surveillance requirements are consistent with the design basis of a

fuel handling accident as documented in the FCS Updated Safety

Analysis Report (USAR).

CONTROL ROOM SYSTEMS

The control room air conditioning (A/C) system consists of two

redundant A/C units, VA-46A and VA-46B. Each unit has sufficient

capacity to meet the cooling requirements for personnel and

equipment inside the control room envelope. Each A/C unit is

equipped with an air-cooled condenser located inside a protective

enclosure outdoors on the roof of the Auxiliary Building. Each A/C

unit's refrigerant compressor, air cooling coils, fans, and dampers

are located inside of the control room envelope. Each unit has a

waterside economizer coil that allows air cooling with Component

Cooling Water (CCW). When cooling water temperature is sufficiently

low, a temperature-activated valve at each A/C unit allows cooling

water flow through the waterside economizer. This valve also diverts

flow away from the waterside economizer if cooling water temperature

is too high. The air-operated CCW isolation valves to the A/C units

fail closed and are automatically closed on a Ventilation Isolation

Actuation Signal (VIAS) to prevent CCW flow through the waterside

economizers in a post-accident situation.

Technical Specification (TS) 2.12(1) requires that the

temperature within the control room and control cabinets be

maintained below 120 deg.F does not meet any of the four criteria

contained in 10 CFR 50.36 for inclusion in TS. However, the

equipment required to maintain this temperature, the control room

air conditioning system, meets Criterion 3 of 10 CFR 50.36 in that

the system functions to mitigate a design basis accident by

maintaining the control room in a habitable environment.

Therefore, it is proposed that this TS be revised to delete the

control room temperature as a LCO and require that two control room

air conditioning trains be operable when the reactor coolant

temperature is above 210 deg.F. The design temperature limits of

instrumentation and controls inside of the control room will be

maintained in the Basis Section of TS 2.12.

The allowed outage time for one train of control room air

conditioning is proposed as 30 days. This is consistent with

Combustion Engineering Standard TS 3.7.12 (NUREG-1432 Rev. 1). In

addition, the FCS Probabilistic Risk Assessment model was reviewed

and validated a 30 day outage time as being non-risk significant.

The impact on Core Damage Frequency (CDF) from a 30 day LCO was

based on the assumption that one cooling unit was always inoperable

and thus under the LCO for an entire year. This allows the analysis

to consider unlimited entries into the LCO and a full LCO duration

for each entry. Using this assumption, the baseline (annually) CDF

of 1.53E-5 would increase by 21.6% to a frequency of 1.86E-5. In

accordance with EPRI's ``PSA Applications Guide,'' this small

increase in CDF can be classified as ``non-risk significant.''

Specification 2.12(2)

Specification 2.12(2) requires that a thermometer be in the

control room at all times. This instrumentation does not meet any of

the four criteria contained in 10 CFR 50.36 for inclusion in the FCS

TS. Therefore, the requirement is proposed for relocation to the FCS

USAR.

Specification 2.12(3)

Specification 2.12(3) requires that all areas of the plant

containing safety related instrumentation be observed during hot

functional testing to determine local temperatures and monitored

during operation if normal plant ventilation is not available. It is

proposed to delete this TS. The requirement to monitor and determine

local temperatures during hot functional testing was met during the

initial startup phase of FCS and is no longer applicable. The

requirement to monitor temperatures within the plant during normal

operation does not meet any of the four criteria contained in 10 CFR

50.36 for inclusion in TS and therefore is being deleted.

The requirement to control temperatures for safety related

instrumentation and controls, and initiate supplementary cooling if

required, is currently described in USAR Section 9.10. These USAR

requirements are controlled by plant procedures. Any changes to

these requirements would require an evaluation be conducted in

accordance with 10 CFR 50.59.

Specification 2.12(4)

Specification 2.12(4) allows one control room air filtration

system to be inoperable for 7 days or a plant shutdown be commenced.

This specification does not state which modes of operation it

applies to.

Therefore, it is proposed to revise this specification to

require two trains of control room air filtration systems to be

operable when the reactor coolant temperature is above 210 deg.F.

The allowed outage time will be maintained at 7 days and a total of

42 hours will be allowed to take the plant to cold shutdown. The 42

hour time period is consistent with TS 2.0.1 which addresses

equipment outages in excess of what is specifically allowed by

individual specifications.

The proposed changes for the control room systems consist of

providing additional restrictions on operation of the control room

air filtration systems and control room air conditioning system.

These changes ensure that equipment required to mitigate the

consequences of an accident are operable. Therefore, the proposed

changes do not increase the probability or consequences of an

accident previously evaluated.

REFUELING OPERATIONS

The design bases of the fuel handling accident and refueling

operations were reviewed and several inadequacies were identified

related to refueling operations. Therefore, revisions are proposed

for the TS Definition section, TS 2.6 on containment integrity, and

TS 2.8 on refueling operations to reflect NUREG-1432.

Definitions

Cold Shutdown Condition & Refueling Shutdown Condition

The changes proposed for the definitions of Cold Shutdown

Condition, and Refueling Shutdown Condition clarify these

definitions. The plant is in Cold Shutdown when Tcold is

less than 210 deg.F, and the reactor coolant is at least Shutdown

Boron Concentration but less than Refueling Boron Concentration.

Similarly, the definition for Refueling Shutdown is clarified to

apply when Tcold is less than 210 deg.F and the reactor

[[Page 30640]]

coolant is at least Refueling Boron Concentration. This change does

not propose any new operating modes but merely clarifies when the

definitions are applicable.

Core Alterations

The definition for Core Alterations is being revised to reflect

the requirements of NUREG-1432. This revision deletes ``any

component'' from the definition and clarifies that the components

considered by this definition are those that could affect

reactivity. In addition, the revision adds nuclear fuel to the

definition such that movement of fuel within the reactor vessel will

be defined as a core alteration and not a refueling operation.

Refueling Operations

The definition of Refueling Operations is being revised to

delete control element assemblies (CEA) or startup sources from the

definition since these are items that are included in the definition

of Core Alterations. Additionally, it is being revised to specify

that the definition is limited to movement of irradiated fuel

outside of the reactor pressure vessel since fuel movement inside

the reactor vessel is included in the definition of Core Alteration.

Finally, a clarification is being added to state that suspension of

refueling operations shall not preclude completion of movement of

irradiated fuel to a safe, conservative position.

In Operation

The definition of In Operation is being revised to include the

definition of operable. This is a more conservative interpretation

than currently exists.

Specification 2.1 ``Reactor Coolant System''

It is proposed to revise TS 2.1.1(3) to include shutdown cooling

requirements when the reactor coolant system (RCS) temperature is

below 210 deg.F with fuel in the reactor and the reactor vessel head

fully tensioned. The definitions of Cold Shutdown (Mode 4) and

Refueling Shutdown (Mode 5) contained in the TS make no distinction

as to the status of the reactor vessel head or RCS temperature. The

only difference between the two defined modes is boron

concentration. Higher or lower boron concentration affects shutdown

margin but does not affect decay heat load, which is the basis for

this specification.

Technical Specification 2.1.1(4) was intended to address

shutdown cooling requirements during refueling operations. However,

this is already addressed in TS 2.8. Therefore, it is proposed to

delete TS 2.1.1(4) and the exception since new specifications

addressing shutdown cooling loop requirements during Mode 5 with

fuel in the reactor and with one or more reactor vessel head closure

bolts less than fully tensioned are proposed for inclusion in TS 2.8

(Refueling Operations).

The associated statements supporting these items in the Basis

section are also proposed for deletion. Prior to any reactor vessel

head closure bolts being loosened, TS 2.1.1 will be applicable which

will require two shutdown cooling loops. As soon as a closure bolt

is loosened, the new proposed TS 2.8 would be applicable which also

requires two shutdown cooling loops whenever there is less than 23

feet of water above the core. The requirements of TS 2.1.1(3) are

similar to NUREG-1432, Specifications 3.4.7 and 3.4.8.

Specification 2.6 ``Containment System''

Currently, TS 2.6(1)c states that containment integrity shall

not be violated when the reactor vessel head is removed if the boron

concentration is less than refueling concentration. However,

Specification 2.6(1)c has no required actions and therefore, TS

2.0.1 must be entered when the LCO is not met. In this situation,

(reactor vessel head removed), TS 2.0.1 is ineffective because the

plant would already be in Refueling Shutdown. Thus, TS 2.6(1)c is

proposed for deletion.

Currently, Specification 2.6(1)d requires that except for

testing one control element drive mechanism at a time, positive

reactivity changes shall not be made by CEA motion or boron dilution

unless containment integrity is intact. Specification 2.6(1)d is

proposed for deletion as it is unnecessarily restrictive.

Specification 2.8.1(1) as proposed eliminates the need for

containment integrity when the reactor is in Refueling Shutdown.

Specification 2.8.1(1) requires sufficient shutdown margin to

preclude a criticality event and also prescribes actions to restore

the shutdown margin if necessary. Small positive reactivity

increases whether by CEA motion or boron dilution will not cause a

criticality event due to the need to maintain at least a 5% shutdown

margin. Therefore, the requirement to maintain containment integrity

is unnecessarily restrictive since a criticality event cannot occur

when a shutdown margin of at least 5% exists. Specification 2.8.1(1)

is consistent with the requirements of NUREG-1432, Specification

3.9.1.

A new specification (TS 2.8.2(1)) is proposed that provides

requirements for containment closure during core alterations and

refueling operations inside of containment. The design basis of the

Fort Calhoun Station does not require full containment integrity

during a fuel handling accident. As stated in USAR Section 14.18,

the fuel handling accident does not take credit for containment

isolation. Therefore, requiring full containment integrity is

inappropriate and requirements for containment closure are proposed

for addition to TS 2.8 consistent with NUREG-1432 Specification

3.9.2.

Specification 2.10.2 governs operation of CEAs and monitoring of

selected core parameters. Specification 2.10.2 ensures (1) adequate

shutdown margin following a reactor trip, (2) that the moderator

temperature coefficient (MTC) is within the limits of the safety

analysis, and (3) CEA operation is within the limits of the setpoint

and safety analysis. Specification 2.10.2 ensures that the reactor

will be maintained sufficiently subcritical to preclude inadvertent

criticality and provides actions (i.e., boration) to be taken to

ensure that the required shutdown margin is available. Thus, TS

2.10.2 precludes the need for containment integrity when the plant

is in cold shutdown.

Specification 2.8 ``Refueling Operations''

It is proposed that TS 2.8 be rewritten to reflect NUREG-1432.

Currently, this specification applies to any refueling operation.

However, no distinction is made between refueling operations within

containment and refueling operations within the spent fuel pool. In

addition, several initial assumptions of a fuel handling accident

are not addressed by the current TS 2.8.

Specification 2.8(1)

The current TS 2.8(1) is inadequate. This specification requires

that the equipment hatch and one door in the Personnel Air Lock be

properly closed, and all automatic containment isolation valves be

operable or at least one valve closed. The specification does not

define what is meant by a properly closed equipment hatch; that

information is currently contained in the Basis of TS 2.1.1. In

addition, inclusion of all automatic containment isolation valves

instead of those providing direct access to the outside atmosphere

is incorrect.

The containment isolation system is defined in USAR Section

5.9.5 as those devices actuated by a Containment Isolation Actuation

Signal (CIAS) or a Steam Generator Isolation Signal (SGIS). This

includes many valves that have no design basis function during a

fuel handling accident. A CIAS is initiated by a Containment

Pressure High Signal or a Pressurizer Pressure Low Signal. Neither

of these signals are required to be operable during refueling

operations as these signals would/could not respond to a fuel

handling accident.

The correct requirements are specified in TS 2.8(2) which only

requires that closure be initiated by the Ventilation Isolation

Actuation Signal (VIAS) for the containment pressure relief, air

sample, and purge system valves. Due to these inadequacies, it is

proposed to delete TS 2.8(1) and replace it with a new Specification

2.8.2(1) which is consistent with NUREG-1432 Specification 3.9.3.

Specification 2.8(2)

It is proposed that TS 2.8(2) be deleted and replaced by new

Specifications 2.8.2(3) and 2.8.3(5). The requirement to maintain an

operable Ventilation Isolation Actuation Signal with input from the

containment atmosphere gaseous and auxiliary building exhaust stack

gaseous radiation monitors is consistent with current requirements

and required actions are consistent with NUREG-1432, Specification

3.3.8. Radiation Monitor RM-052 functions as a ``swing'' monitor,

i.e., it can be aligned to monitor either containment or the

auxiliary building exhaust ventilation stack. Radiation Monitor RM-

052 is powered by either MCC-3B1/AI-40C (like RM-051) or MCC-4C2/AI-

40D (like RM-062).

Technical Specification 2.7, Electrical System is not required

to be applied when the RCS is below 300 deg.F. Above 300 deg.F, TS

2.7 requires both 4160-VAC buses to be operable. Thus, above

300 deg.F the required radiation monitors must be powered from

independent 480-VAC buses supplied by independent 4160-VAC buses.

During refueling outages, bus alignments other than those used

during power operation are used to permit electrical system

maintenance and modifications.

In the loss of offsite power event, the radiation monitor sample

pumps and control room HVAC units stop and will not restart

[[Page 30641]]

until the emergency diesel generators (EDGs) reenergize the system.

The fuel handling equipment also stops and does not restart when the

EDGs reenergize the system, thus minimizing the potential of a fuel

handling accident. When the EDGs reenergize the buses, VIAS will

operate as designed. Therefore, when the RCS is below 300 deg.F, the

required monitors need only be powered from independent 480-VAC

buses supplied by a single 4160-VAC bus.

There is no need to assume that a fuel handling accident occurs

immediately followed by a loss of offsite power. However, in the

unlikely event that this should occur, there would be no effect on

the site boundary dose since VIAS is not credited in USAR Section

14.18 (Fuel Handling Accident). In this situation, when the EDGs

reenergize the buses, the control room HVAC units will restart in

the filtered air makeup mode and the stack radiation monitor sample

pump will restart. However, the containment radiation monitor sample

lines remain isolated preventing the restart of the monitor sample

pump after receipt of a VIAS.

Specification 2.8(3)

It is proposed that TS 2.8(3) be deleted. This requirement does

not meet any of the four criteria contained in 10 CFR 50.36 for

inclusion in the TS. The requirement that radiation levels in

containment and the spent fuel pool shall be monitored during

refueling operations will be incorporated into the FCS USAR.

Specification 2.8(6)

It is proposed that TS 2.8(6) be deleted. This requirement does

not meet any of the four criteria contained in 10 CFR 50.36 for

inclusion in the TS. The requirements that direct communication

between personnel in the control room and at the refueling machine

shall be available whenever core alterations are taking place will

be incorporated into the FCS USAR.

Specification 2.8(7)

It is proposed that TS 2.8(7) be deleted and replaced with a new

Specification 2.8.3(4). The requirement to place the spent fuel pool

ventilation system in operation prior to refueling operations is

consistent with the current TS. It is being clarified that this

specification only applies to refueling operations in the spent fuel

pool, and not when conducting refueling operations inside of

containment. Additionally, it is being clarified that TS 2.0.1 is

not applicable to this activity, as reactor operation is independent

of fuel movements in the spent fuel pool.

Specification 2.8(9)

The current Specification 2.8(9) is inadequate. This

specification requires a minimum of 23 feet of water above the top

of the core. This does not meet the initial conditions assumed in

the fuel handling accident as documented in USAR Section 14.18. USAR

Section 14.18 assumes 23 feet of water above where the fuel could

land if dropped. In order to meet this initial condition, a minimum

of 23 feet of water above the reactor vessel flange is required, as

this is the highest point where a fuel bundle could land if dropped.

Procedures reflect the requirement to maintain 23 feet of water

above the reactor vessel flange during refueling operations. The

proposed revision is consistent with NUREG-1432, Specification

3.7.16.

Specification 2.8(11)

The current specification is inadequate. The specification

provides restrictions on storage of fuel in the spent fuel pool;

however, there are no required actions to address situations when

the specification is not met. It is proposed that TS 2.8(11) be

deleted and replaced with a new Specification 2.8.3(1) that requires

that a misloaded fuel assembly be moved immediately. Additionally,

it is being clarified that TS 2.0.1 is not applicable to this

activity, as reactor operation is independent of fuel movements in

the spent fuel pool.

Specification 2.8(12)

It is proposed that TS 2.8(12) be deleted and replaced with a

new Specification 2.8.3(3). The requirement to maintain 500 ppm

boron concentration in the spent fuel pool whenever unirradiated

fuel is stored there is consistent with the current TS and the

required actions are consistent with NUREG-1432, Specification

3.7.17.

Restriction on Movement of Irradiated Fuel from the Reactor Core

The restriction on irradiated fuel movement unless the core has

been subcritical for at least 72 hours if the reactor has been

operated at power levels above 2% is proposed for relocation to the

Bases of TS 2.8.2(2). This requirement does not meet any of the four

criteria contained in 10 CFR 50.36 for inclusion in the TS. This is

consistent with NUREG-1432, B 3.9.6.

Reactor Coolant System Boron Concentration

Currently, there is no specification for boron concentration.

Refueling boron concentration is included in the definition of Mode

5. However, there are no required actions to be taken if the boron

concentration should be below refueling concentration. Therefore, it

is proposed that a new Specification 2.8.1(1) be incorporated

consistent with NUREG-1432, Specification 3.9.1.

Spent Fuel Pool Water Level

Currently, there is no specification for spent fuel pool water

level. The water level of the spent fuel pool is an initial

condition assumed in USAR Section 14.18. It is proposed that a new

Specification 2.8.3(2) be incorporated into TS 2.8, which is

consistent with NUREG-1432, Specification 3.7.16.

The proposed changes for the RCS and containment during

shutdown, and requirements for refueling operations, consist of

providing additional restrictions on operation, and changes to make

the requirements of the TS Limiting Conditions for Operation

consistent with the initial conditions and assumptions of the fuel

handling accident as documented in USAR Section 14.18. Therefore,

the proposed changes do not increase the probability or consequences

of an accident previously evaluated.

SURVEILLANCE REQUIREMENTS

CONTROL ROOM

Specification 3.1, Table 3-3, Item 13.

Specification 3.1, Table 3-3, Item 13 requires that the

thermometer in the control room be compared with a calibrated

thermometer and replaced if out of tolerance on a refueling

frequency. It is proposed that this surveillance be deleted to

be consistent with deletion of the LCO requirement to maintain a

thermometer in the control room.

A new surveillance is proposed to verify that the control room

air conditioning system has the capability to remove the assumed

heat load. This surveillance will ensure the operability

requirements for TS 2.12 are met. The test and frequency is

consistent with NUREG-1432.

The air-operated CCW isolation valves to the A/C units fail

closed and are automatically closed on a VIAS to prevent CCW flow

through the waterside economizers in a post-accident situation.

These valves are currently tested in accordance with TS 3.3 (FCS

Inservice Testing Program). Prior to the modification, the valves

were tested as fail-open valves. No TS changes are necessary.

The control room air filtration system is currently tested on a

refueling frequency in accordance with TS 3.2, Table 3-5, Item 10a.

No TS changes are necessary.

REFUELING OPERATIONS

Reactor Coolant Boron Concentration During Refueling Operations

The Reactor Coolant System boron concentration is currently

sampled in accordance with TS 3.2, Table 3-4, Item 1(e). It is

proposed to revise the frequency from once per shift during

refueling operations to once per 3 days which is consistent with

NUREG-1432. As stated in the basis of TS 2.8 and USAR Section 14.18,

the reactor cavity is filled with over 200,000 gallons of borated

water prior to the start of refueling operations. The requirements

for sampling the reactor coolant during the remainder of Mode 5 is

performed once per 3 days in accordance with Table 3-4, Item 1(d).

This proposed change will make the sampling consistent with the

requirements of Item 1(d) and NUREG-1432.

Spent Fuel Pool Boron Concentration

The spent fuel pool boron concentration is currently sampled in

accordance with TS 3.2, Table 3-4, Item 5. It is proposed to revise

the frequency of the sampling to prior to movement of unirradiated

fuel in the spent fuel pool and once per week whenever unirradiated

fuel is stored there to be consistent with the requirements of the

LCO.

Source Range Neutron Monitors

Currently, a channel check and calibration of the wide range

neutron monitors is performed in accordance with TS 3.1, Table 3-1,

Item 2.

Containment Penetrations

Currently, there is no surveillance to determine the status of

containment penetrations during refueling operations. Therefore, a

new surveillance is proposed for TS 3.2, Table 3-5 to verify the

status of required containment penetrations once per 7 days

consistent with NUREG-1432.

The requirement of NUREG-1432 to verify that the containment

purge and exhaust valves actuate to the isolation position on a

refueling frequency is currently tested as part of the Containment

Radiation High Signal test required by TS 3.1, Table 3-2. Item 4.

Shutdown Cooling Loops

Currently, there is no surveillance requirement to verify that

the required

[[Page 30642]]

shutdown cooling loops are operable and in operation or to verify

correct breaker lineup for the shutdown cooling loop that is not in

operation. Therefore a new surveillance is proposed to be

incorporated into TS 3.2, Table 3-5 consistent with NUREG-1432.

Refueling Water Level

Currently, there is no surveillance requirement to verify the

refueling water level during refueling operations. Therefore, a new

surveillance is proposed for incorporation into TS 3.2, Table 3-5

consistent with NUREG-1432.

Spent Fuel Pool Water Level

Currently, there is no surveillance requirement to verify the

spent fuel pool water level during refueling operations. Therefore,

a new surveillance is proposed for incorporation into TS 3.2, Table

3-5 consistent with NUREG-1432.

Spent Fuel Initial Enrichment/Burnup Verification

Currently, the requirement to conduct a verification of initial

enrichment and burnup of spent fuel that will be stored in Region 2

is included as a general requirement of TS 2.8. It is proposed to

relocate this requirement into a surveillance in TS 3.2, Table 3-5,

consistent with NUREG-1432.

The proposed changes for the surveillance requirements consist

of providing additional testing requirements to ensure that the

Limiting Condition for Operations will be met. One surveillance

frequency related to the sampling of the reactor coolant system

boron concentration during refueling operations is being reduced

from a frequency of once per shift to once every 3 days. However,

this frequency is consistent with the frequency of sampling during

the remainder of Mode 5 when fuel is in the

reactor and is more than adequate due to the large volume (over

200,000 gallons) of borated water required during refueling

operations. Therefore, the proposed changes do not increase the

probability or consequences of an accident previously evaluated.

ADMINISTRATIVE CHANGES

The remainder of TS 2.8 requirements of refueling operations are

proposed to be reformatted into individual TS LCOs. It is also

proposed that sampling frequencies of items contained in TS 3.2,

Table 3-4, (page 3-19), be revised to incorporate frequencies

defined in TS 3.0.2. Therefore, frequencies stated as once per 31

days will be noted as ``M,'' and frequencies stated as once per 7

days will be noted as ``W.'' These proposed changes have no effect

on the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

There will be no physical alterations to the plant

configuration. No changes in operating modes are proposed although

minor changes to the definitions of Cold Shutdown Condition and

Refueling Shutdown Condition are proposed for clarification

purposes. The proposed changes incorporate additional restrictions

on the operation and testing of equipment required to mitigate an

accident and to ensure the initial conditions and assumptions of the

design basis accidents are maintained and controlled by the

Technical Specifications.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes ensure that assumptions of the fuel

handling accident are maintained by Technical Specification Limiting

Condition for Operation and surveillance requirements. The

assumptions of the fuel handling accident that may affect a margin

of safety are not being changed. Therefore, the proposed change does

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

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

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

Pennsylvania

Date of amendment request: February 25, 1997

Description of amendment request: The proposed Technical

Specifications (TS) changes would amend the Limerick Generating Station

(LGS) Unit 1 and Unit 2 Facility Operating Licenses (FOLs), and

Appendix B of the licenses (i.e., Environmental Protection Plan (EPP)),

reflecting a corporate name change from Philadelphia Electric Company

to PECO Energy Company. In addition, the application would make changes

to the LGS Units 1 and 2, FOL, and Appendix A (i.e., TS) of the

licenses, which would remove obsolete information and correct

typographical errors.

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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The company name change and typographical corrections are

editorial and will not alter the operation of equipment assumed to

be an initiator of any analyzed event or transients previously

evaluated. The license provisions were satisfactorily completed, and

as such, have no effect on any previously evaluated accident

scenario. The changes will not alter the operation of equipment

assumed to be available for the mitigation of accidents or

transients, nor will they alter the operation of equipment important

to safety previously evaluated.

Therefore, the changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The company name change and typographical corrections are

editorial and will not involve any physical changes to the plant

systems, structures, or components. The license provisions were

satisfactorily completed, and as such, have no effect on any

previously evaluated accident scenario. The proposed changes do not

allow plant operation in any mode that is not already evaluated. The

changes will not alter the operation of equipment important to

safety previously evaluated.

Therefore, the changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The company name change and typographical corrections are

editorial and will not affect the manner in which the facility is

operated, or change equipment or features which affect the

operational characteristics of the facility. There is no margin of

safety as defined in the bases of any TS regarding the name of the

company, or affected by the corrections or deletion of obsolete

license provisions.

Therefore, these 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

[[Page 30643]]

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

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

Pennsylvania

Date of amendment request: March 24, 1997

Description of amendment request: The proposed Technical

specifications (TS) changes would delete the Drywell and Suppression

Chamber Purge System operational time limit, and add a surveillance

requirement to ensure the purge system large supply and exhaust valves

are closed as required.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

This activity does not increase the probability of occurrence of

an accident previously evaluated in the SAR [Safety Analysis

Report]. This activity involves deleting the allowable operating

limit (180 hours each 365 days) for the Drywell and Suppression

Chamber Purge system, while maintaining specific criteria for when

the valves are allowed to be open. These changes do not increase the

probability that this system will be in service should a LOCA [loss-

of-coolant-accident] occur and does not increase the probability

that a LOCA will occur. These changes also do not impact the

probability of occurrence of any anticipated operational occurrence,

other postulated design basis accident, or other event in which the

plant was designed to respond.

This activity does not increase the consequences of an accident

previously evaluated in the SAR. UFSAR [Updated Final Safety

Analysis Report] Section 9.4.5.1.2.2 for high volume purging,

although limiting the operating time the vent and purge system is to

be in service, evaluates the consequences of a LOCA should the vent

and purge valves be open. System operating procedures for venting

and purging assure the availability of SGTS [standby gas treatment

system] should a LOCA occur.

This activity will not increase the probability of a LOCA

occurring during the time the Drywell and Suppression Chamber Purge

system is in operation as previously evaluated. The Improved TS do

not identify a specific time limit value as long as the valves are

operated under the stated conditions (inerting, de inerting,

pressure control, ALARA [as low as reasonably achievable] or air

quality considerations for personnel entry or Surveillances that

require that the valves be open). These proposed changes will

incorporate the ITS [Improved Technical Specifications] operational

controls which will result in the same order of magnitude of

equipment malfunction probability as that provided by limiting

purging to 180 hours per 365 days. A LGS [Limerick Generating

Station] Level 2 PSA [Probability Risk Assessment] Analysis was

performed to determine the additional risk associated with changing

the operating limit from 90 hours to a nominal 500 hours each 365

days. This analysis concluded that the increase in risk of

containment failure is well within the bounds of the EPRI [Electric

Power Research Institute] PSA Applications Guideline for permanent

changes and the NRC [Nuclear Regulatory Commission] Staff's safety

goal value of 1.0 E-6 per year of reactor operation. Industry and

LGS historical operating experience confirms that the purging lines

are opened only for the specified reasons stated in ITS and for

periods which do not exceed the current magnitude of equipment

malfunction probability. Therefore, earlier engineering judgment is

being replaced by operating experience.

Failure of the operating SGTS filter bank following a LOCA has

been found to be acceptable due to the limited benefit derived from

SGTS for accident sequences important to plant risk and the

possibility that the backup filter bank would be available.

Additionally, as discussed in UFSAR Section 9.4.5.1.2.2, the failure

of SGTS during a LOCA does not contribute to any significant

releases and is bounded by the analysis performed to address

containment overpressure rupture.

Deleting the time limit restriction that the vent and purge line

isolation valves may be open does not increase the probability that

these valves will not perform as designed (close upon isolation

signal) in response to a LOCA. Removing the 180 hour requirement

will not increase the likelihood that the vent and purge valves will

be called upon to close from that previously evaluated. UFSAR

Section 6.2 states that the containment purge valves have undergone

extensive testing and analyses to demonstrate the operability of

these valves following a LOCA.

These changes do not directly or indirectly degrade the

performance of any other safety system (assumed to function in the

accident analysis) design basis. The potential for other equipment

failures in the reactor enclosure due to duct impact, impingement,

and the resulting environmental conditions was previously evaluated

in the LGS SAR. It was concluded that the environmental

qualifications for the LGS equipment are sufficient to ensure

operability under the predicted environmental condition, and, the

potential does not exist for impact or impingement - related damage

to essential equipment. Maintaining the existing SAR analysis and

retaining operating criteria for opening the containment purge

valves, demonstrates that the risk of equipment failure and

resulting radiological consequences will not increase.

Therefore, deleting the TS operating limit for the Drywell and

Suppression Chamber Purge system from 180 hours each 365 days and

the addition of a TS Surveillance Requirement verifying that the

purge valves are closed under certain conditions does not increase

the probability or consequences of an accident previously evaluated.

2. The proposed Technical Specifications changes do not create

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

accident previously evaluated.

This activity does not change the function of the Drywell and

Suppression Chamber Purge system, the containment isolation system,

or SGTS as previously evaluated. Deleting the operational time limit

that the vent and purge system is in service and the addition of a

surveillance requirement does not create an accident initiator not

already considered.

In addition, this activity does not create a failure mode not

considered. All evaluated equipment failures that could occur as a

result of a LOCA during high volume purging have previously been

identified and evaluated. Therefore, these changes do not create the

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

previously evaluated.

3. The proposed Technical Specifications changes do not involve

a significant reduction in a margin of safety.

The bases of TS 3.6.1.8 state that the 180 hour each 365 day

operating limit for the Drywell and Suppression Chamber Purge system

is imposed to protect the integrity of the SGTS filters. The LGS

Offsite Dose Calculation Manual assures the availability of the

backup SGTS filter train during operation of the vent and purge

system. Furthermore, deleting the operating limit (180 hours each

365 days) does not reduce the margin of safety since specific

criteria for opening the purge valves is being maintained and does

not involve an increase in risk. 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: April 9, 1997

Description of amendment request: The proposed Technical

Specifications (TS) changes would clarify existing battery specific

gravity requirements, delete the requirement to correct specific

gravity values based on electrolyte level, and allow the use of

charging current measurements to verify the batterys state of charge.

Basis for proposed no significant hazards consideration

determination:

[[Page 30644]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration, which is

presented below:

1. The proposed Technical Specifications changes do not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

Changes to Technical Specifications surveillance requirements

for specific gravity and Technical Specifications Bases commitments

do not change the frequency or consequences of any accident

previously evaluated. The proposed changes which commit to IEEE

Standard 450-1995 for specific gravity testing, providing battery

charging current as an alternate method to specific gravity

measurements, and eliminating the commitment to perform electrolyte

level correction do not prevent the DC system from performing its

intended safety function. The proposed changes to the Technical

Specification battery surveillance requirements and commitment to

IEEE Standard 450-1995 for specific gravity are in accordance with

current industry practices. These changes do not reduce the

readiness and performance of the 1E DC power system to perform its

intended function during a design basis event.

The proposed changes do not affect seismic specifications,

separation criteria or environmental qualifications. The proposed

changes do not impose an increase in or more severe test

requirements, an increase in the frequency of operation, reduce

independence or redundancy, modify the system or equipment

protective features, introduce new equipment failures or impose

additional loads than any previously evaluated. The Class 1E battery

system will continue to meet all of the design standards applicable

to the system and will not cause the system to operate outside of

its design or testing limits.

Batteries or battery chargers and their failure are not

initiators of the accidents previously evaluated. The proposed

changes do not affect, degrade or prevent the response of active or

passive systems described or assumed in the LGS accidents previously

evaluated. In addition, the proposed TS changes will improve the

availability of the station batteries.

Therefore, the changes will not increase the probability or

consequences of an accident previously evaluated.

2. The proposed Technical Specifications changes do not create

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

accident previously evaluated.

The proposed Technical Specifications changes which will revise

the surveillance requirements and the TS Bases, do not increase the

failure rate of the battery. The proposed changes clarify and

enhance Operation's focus on the key battery parameters which will

improve the availability of the station batteries. The station

batteries are not accident initiators. The single failure of an

electrical component was previously evaluated in the LGS accident

analysis. Unexpected failures beyond the postulated single failure

are no more likely to occur under the clarified surveillance

requirements.

Therefore, these changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

3. The proposed Technical Specifications changes do not involve

a significant reduction in a margin of safety.

The revision clarifies and reduces the battery surveillance

requirements for specific gravity. The revision eliminates the

possibility for misinterpretation and provides consistency of the

surveillance requirements. The specific gravity value for each

connected cell is being revised to reflect a discrete number which

meets the existing manufacturer's recommendations and does not

differ from the value described in the present bases. LGS is

currently committed to earlier revisions of IEEE Standard 450 (i.e.,

1975 and 1980), and the incorporation of IEEE Standard 450-1995 for

specific gravity will reflect current industry practices regarding

specific gravity.

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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear

Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: April 30, 1997 (TS 97-01)

Description of amendment request: The proposed amendment would

change the design features section of the Technical Specifications to

provide for insertion of Lead Test Assemblies (LTAs) containing Tritium

Producing Burnable Absorber Rods (TPBARs) in the Watts Bar Nuclear

Plant (WBN) reactor core during Cycle 2. The purpose of the change is

to provide irradiation services to support U.S. Department of Energy

(DOE) investigations into the feasibility of using commercial light

water reactors to maintain the DOE inventory of tritium.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

LTAs do not adversely affect reactor neutronic or thermal-

hydraulic performance; therefore, they do not significantly increase

the probability of accidents or equipment malfunctions while in the

reactor. The neutronic behavior of the LTAs mimics that of standard

burnable absorbers with only slight differences which are

accommodated in the core design. The reload safety analysis

performed for WBN Unit 1, Cycle 2 will confirm that any minor

effects of LTAs on the reload core will be within established fuel

design limits.

As described in DOE Technical Report PNNL-11419, Revision 1, the

LTA design is robust to all accident conditions except the large

loss of coolant accident where the rods are susceptible to failure.

However, the failure of the small number of TPBARS rods has been

determined to have an insignificant effect on the thermal hydraulic

response of the core to this event.

The impacts of LTAs on the radiological consequences for certain

postulated events [as shown in Table 6-1 of the licensee's

submittal, including Large Break LOCAs{time} are very small, and

they remain within 10 CFR 100 regulatory limits. The additional

offsite doses due to tritium leakage from the containment are small

with respect to loss of coolant accident source terms and are well

within regulatory limits.

The LTAs will not result in an increase in combustible gas

released to the containment. Therefore, the LTAs do not result in a

significant increase in the consequences of those previously

considered.

Analysis has shown that TPBARs will not fail during Condition I

through IV events, with the exception of a Large Break LOCA. The

radiological consequences of the non-Large-Break LOCA events are

essentially unchanged by the expected TPBAR tritium leakage to

reactor coolant, and doses remain within a small fraction of 10 CFR

100 regulatory limits. Therefore, there is no significant increase

in the consequences of these previously evaluated accidents.

The expected occupational and offsite doses, as reported in

Technical report PNNL-11419, Revision 1, resulting from release of

tritium from TPBARs over the plant operating cycle, including

refueling, are not significantly increased and are within applicable

regulatory limits.

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

new or different kind of accident from any accident previously

evaluated.

LTAs have been designed to be compatible with existing

Westinghouse 17x17 fuel assemblies and conventional Burnable Poison

Rod Assembly (BPRA) handling tools, equipment, and procedures, and

therefore no new accidents or equipment malfunctions are created by

the handling of LTAs.

LTAs use materials with known and predictable performance

characteristics and are compatible with PWR coolant. The LTA design

has specifically included material similar to those used in standard

burnable absorber rods with the exception of internal

[[Page 30645]]

assemblies used in the production and retention of tritium. As

described in the technical report, these materials are compatible

with the reactor coolant system and the core design. For the

irradiation proposed, the quantities of these materials is small.

Therefore, no new accidents or equipment malfunctions are created by

the presence of the LTAs in the reactor coolant system.

Thermal-hydraulic criteria have been established to ensure that

TPBARs will not fail during Condition I or II events. Analysis has

shown that TPBARs, appropriately positioned in the core, operate

within the established thermal-hydraulic criteria. Therefore, no new

accidents or equipment malfunctions are created by the presence of

the LTAs in the reactor.

Analysis has shown that TPBARs will not fail during Condition

III and IV events, with the exception of a large-break loss-of-

coolant-accident. The radiological consequences of these events are

small, with doses that are a small fraction of the 10 CFR 100

limits. Therefore there is no significant increase in consequences

of these previously evaluated accidents.

LTAs do not adversely affect reactor neutronic or thermal-

hydraulic performance; therefore, they do not create the possibility

of accidents or equipment malfunctions of a different type than

previously evaluated while in the reactor.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

LTAs do not adversely affect reactor neutronic or thermal-

hydraulic performance. Analysis indicates that reactor core behavior

and offsite doses remain relatively unchanged. TPBAR performance

under Condition I, II, III, and IV events are very similar to

standard burnable absorber rods previously evaluated. For these

reasons, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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: April 18, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 3/4.3.2, ``Safety System

Instrumentation,'' and TS Section 3/4.5.2, ``Emergency Core Cooling

Systems - ECCS Subsystems - Tavg (greater than or equal to)

280 deg.F.'' Certain surveillance intervals would be changed from 18

months to once each refueling interval, and certain setpoints would be

changed. The associated bases would also be changed.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

The Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) has

reviewed the proposed changes and determined that a significant

hazards consideration does not exist because operation of the DBNPS,

in accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because the initiation of such

accidents are not affected by the proposed revisions to increase the

surveillance test intervals from 18 to 24 months for TS 3/4.3.2.1,

``Safety Features Actuation System Instrumentation,'' and TS 3/

4.5.2, ``Emergency Core Cooling Systems - ECCS Subsystems - Tavg

(greater than or equal to) 280F.'' Initiating conditions and

assumptions remain as previously analyzed for accidents in the DBNPS

Updated Safety Analysis Report.

Results of the instrument drift study analysis and review of

historical 18-month surveillance data and applicable maintenance

records support an increase in the surveillance test intervals from

18 to 24 months (and up to 30 months on a non-routine basis)

because: the projected instrument errors caused by drift are bounded

by the existing setpoint analysis or a new analysis has been

performed incorporating a more conservative setpoint; and no

potential for a significant increase in a failure rate of a system

or component was identified during surveillance data and applicable

maintenance records reviews.

These proposed revisions are consistent with the NRC guidance on

evaluating and proposing such revisions as provided in Generic

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

Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2,

1991.

The proposed revisions to Allowable Values for Safety Features

Actuation System (SFAS) Reactor Coolant System (RCS) Pressure - Low,

RCS Pressure - Low-Low, RCS Pressure - Low-Low bypass permissive,

and Decay Heat Isolation Valve and Pressurizer Heater Interlocks

have no bearing on the probability of the initiation of an accident

previously evaluated.

The application of the Allowable Value to only the Channel

Functional Test and not the Channel Calibration, the proposed

deletion of the Trip Setpoints, the proposed revision of the TS

3.3.2.1 Limiting Condition for Operation (LCO) and Action Statement

3.3.2.1.a, and the proposed revisions to Actions 13 and 14 of TS

Table 3.3-3, are associated with the proposed revision of the

Allowable Values for SFAS RCS Pressure - Low, RCS Pressure - Low-

Low, and Decay Heat Isolation Valve and Pressurizer Heater

Interlocks, and are consistent with NUREG-1430, Revision 1,

``Standard Technical Specifications, Babcock and Wilcox Plants,''

dated April 1995. The proposed revisions have no bearing on the

probability of the initiation of an accident previously evaluated.

The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor

Protection System and Safety System Instrumentation,'' and TS Bases

3/4.5.2 and 3/4.5.3, ``ECCS Subsystems,'' are administrative changes

associated with the other proposed changes, and do not affect

previously analyzed accidents.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because the slight increase in doses

due to a letdown line break event as a result of the proposed change

to the SFAS RCS Pressure - Low Allowable Value still satisfy the NRC

Standard Review Plan Section 15.6.2 acceptance criteria that doses

do not exceed a small fraction (10%) of the 10 CFR 100 guideline

values. The remaining proposed changes to Allowable Values, and the

other changes proposed by this License Amendment Request do not

increase the radiological consequences of previously analyzed

accidents because the source term, containment isolation, or

radiological releases are not being changed by the proposed

revisions.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated, for the reasons

discussed below.

No changes are being proposed to the type of testing currently

being performed, only to the length of the surveillance test

interval.

Results of the instrument drift study analysis and review of

historical 18-month surveillance data and maintenance records

support an increase in the surveillance test intervals from 18 to 24

months (and up to 30 months on a non-routine basis) because: the

projected instrument errors caused by drift are bounded by the

existing setpoint analysis or a new analysis has been performed

incorporating a more conservative setpoint; and no potential for a

significant increase in a failure rate of a system or component was

identified during surveillance data and applicable maintenance

records reviews.

The proposed revisions to Allowable Values for SFAS RCS Pressure

- Low, RCS Pressure - Low-Low, RCS Pressure Low-Low bypass

permissive, and Decay Heat Isolation Valve and Pressurizer Heater

Interlocks, do not alter the type of any testing currently being

performed.

The application of the Allowable Value to only the Channel

Functional Test and not the Channel Calibration, the proposed

deletion of the Trip Setpoints, revision of the TS 3.3.2.1 LCO and

Action Statement 3.3.2.1.a, and the proposed revisions to Actions 13

and 14 of

[[Page 30646]]

TS Table 3.3-3, are associated with the proposed revision to the

Allowable Values for SFAS RCS Pressure - Low, RCS Pressure - Low-

Low, RCS Pressure Low-Low bypass permissive, and Decay Heat

Isolation Valve and Pressurizer Heater Interlocks, and are

consistent with NUREG-1430, Revision 1, ``Standard Technical

Specifications, Babcock and Wilcox Plants,'' dated April 1995. The

proposed revisions do not alter the type of testing currently being

performed.

The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor

Protection System and Safety System Instrumentation,'' and TS Bases

3/4.5.2 and 3/4.5.3, ``ECCS Subsystems,'' are administrative changes

associated with the other proposed changes, and do not alter any

testing currently being performed.

3. Not involve a significant reduction in a margin of safety.

The results of the instrument drift study analysis and review of

historical 18-month surveillance data and applicable maintenance

records support an increase in the surveillance test intervals from

18 to 24 months (and up to 30 months on a non-routine basis)

because: the projected instrument errors caused by drift are bounded

by the existing setpoint analysis or a new analysis has been

performed incorporating a more conservative setpoint; and no

potential for a significant increase in a failure rate of a system

or component was identified during surveillance data and applicable

maintenance records reviews. Existing system and component

redundancy is not affected by these proposed changes.

There are no new or significant changes to the initial

conditions contributing to accident severity or consequences,

consequently there are no significant reductions in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensees: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: Gail H. Marcus

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: April 18, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 3/4.7.6, ``Plant Systems -

Control Room Emergency Ventilation System.'' Additional Limiting

Conditions for Operation (LCO) would be added related to the

availability of the station vent normal range radiation monitoring

instrumentation. The associated TS bases would also be modified

consistent with these changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

The Davis-Besse Nuclear Power Station has reviewed the proposed

changes and determined that a significant hazards consideration does

not exist because operation of the Davis-Besse Nuclear Power Station

(DBNPS), Unit No. 1, in accordance with this change would not:

1a. 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.

The proposed change to LCO 3.7.6.1 would include new required

Action statements in the event that one or both channels of station

vent normal range radiation monitoring instrumentation become

inoperable. In the event that one channel is inoperable for greater

than 7 days, or in the event that both channels are inoperable, the

proposed Action statement would require that the control room normal

ventilation system be isolated and at least one Control Room

Emergency Ventilation System (CREVS) train be placed in operation.

Under the proposed actions, the ventilation systems would be

placed in a state equivalent to that which occurs were a high

radiation isolation to occur. These proposed changes have no bearing

on the probability of an accident.

The proposed change to the terminology utilized in Surveillance

Requirement (SR) 4.7.6.1.e is an administrative change made to make

the terminology consistent with the proposed new Action statements.

The proposed changes to Bases 3/4.7.6 are administrative changes

consistent with the proposed changes to LCO 3.7.6.1. These changes

have no bearing on the probability of an accident.

1b. Involve a significant increase in the consequences of an

accident previously evaluated because the proposed changes do not

change the source term, containment isolation, or allowable

releases.

As described above, under the proposed new LCO 3.7.6.1 Actions,

in the event that one station vent normal range radiation monitoring

instrumentation channel is inoperable for greater than 7 days, or in

the event that both channels are inoperable, the ventilation systems

would be placed in a state equivalent to that which occurs were a

high radiation isolation to occur. Therefore, in the unlikely event

of an accident requiring control room isolation while in this

condition, the dose consequences to control room operators would be

unchanged.

The proposed change to the terminology utilized in SR 4.7.6.1.e

is an administrative change made to make the terminology consistent

with the proposed new Action statements. The proposed changes to

Bases 3/4.7.6 are administrative changes consistent with the

proposed changes to LCO 3.7.6.1. These changes have no bearing on

the consequences of an accident.

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

from any accident previously evaluated because no new accident

initiators or assumptions are introduced by the proposed changes.

As described above, under the proposed new LCO 3.7.6.1 Actions,

in the event that one station vent normal range radiation

monitoring instrumentation channel is inoperable for greater

than 7 days, or in the event that both channels are inoperable,

theventilation systems would be placed in a state equivalent to that

which occurs were a high radiation isolation to occur. Operation of

the equipment and components in this manner would not introduce the

possibility of any new or different kinds of accidents.

The proposed change to the terminology utilized in SR 4.7.6.1.e

is an administrative change made to make the terminology consistent

with the proposed new Action statements. The proposed changes to

Bases 3/4.7.6 are administrative changes consistent with the

proposed changes to LCO 3.7.6.1. These changes would not introduce

the possibility of any new or different kinds of accidents.

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

the proposed changes to the Action under LCO 3.7.6.1 ensure that

control room isolation capability is maintained in the event a

station vent radiation monitor is inoperable. The proposed allowable

outage time of 7 days for one inoperable channel is consistent with

the presently allowable outage time for one inoperable CREVS. The

proposed Action to place at least one CREVS train in operation

within 1 hour, in the event both channels of radiation monitoring

become inoperable, is more conservative than the present Action

which requires that a plant shutdown commence within 1 hour, but

does not require the CREVS be placed in operation.

The proposed change to the terminology utilized in SR 4.7.6.1.e

is an administrative change made to make the terminology consistent

with the proposed new Action statements. The proposed changes to

Bases 3/4.7.6 are administrative changes consistent with the

proposed changes to LCO 3.7.6.1. These changes would not affect the

margin of safety.The NRC staff has reviewed the licensees' analysis

and, based on this review, it appears that the three standards of 10

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

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

Basis for proposed no significant hazards consideration

determination:

[[Page 30647]]

Attorney for licensees: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: August 22, 1996

Description of amendment request: The proposed change would remove

the action statement of Technical Specification (TS) Section 3.2.G,

Table 3.2.6, Note 7, requiring reactor shutdown after 30 days of

inoperability of the high range stack gas monitor and substitute an

action statement consistent with the guidance provided in NRC Generic

Letter 83-36.

The high range stack monitor provides an estimate of gross stack

activity that has exceeded the upper limit of the normal range

instrumentation. The high range monitor reading serves as input to dose

projection systems for initial estimation of off-site conditions. The

monitor reading would be used prior to the acquisition of stack

isotopic sample data which would provide a more accurate indication of

stack activity.

The licensee stated, among other things, that due to the

passivefunction of the instruments and the ability to monitor this

parameter utilizing alternate methods, it is not appropriate to impose

stringentrequirements on the operation of the unit. This monitor is

identified in the Vermont Yankee Regulatory Guide 1.97 submittal as

Category 2, Type E. This monitor provides post-accident information for

use in determining the magnitude of the release of radioactive

materials and for monitoring such release. However, the high range

stack monitor does not have any safety function associated with the

prevention or automatic mitigation of design-basis accidents, neither

does it provide primary information needed to permit the control room

operating personnel to take required manually controlled actions.

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 TS change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.]

The High Range Stack Monitor is a RG [Regulatory Guide] 1.97,

Category 2, Type E instrument with no specified safety function

associated with the prevention or automatic mitigation of design

basis accidents, neither does it provide primary information needed

to permit the control room operating personnel to take required

manually controlled actions. The proposed change to the action

statement associated with this monitor will not change the function

of this monitor, and since the monitor is not assumed to initiate

any accidents, nor function to mitigate any accidents, this change

will not significantly increase the probability or consequences of

any previously analyzed accident.

[(2) The proposed TS change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.]

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The proposed change will still ensure effective methods

are available to assess post accident conditions. Thus, this change

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

accident from any accident previously evaluated.

[(3) The proposed TS change does not involve a significant

reduction in a margin of safety.]

The proposed change to the action statement associated with

this monitor will not change the function of this monitor, and since

the monitor is not assumed to function for the prevention or

mitigation of any previously evaluated accidents, this change will

not significantly reduce 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Attorney for licensee: R. K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624

NRC Project Director: Patrick D. Milano, Acting

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

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

Creeks, Manitowoc County, Wisconsin

Date of amendment request: January 24, 1997, as supplemented on May

15, 1997 (TSCR 193)

Description of amendment request: The proposed amendments (Point

Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request

(TSCR) 193) would revise TS 15.5.4, ``Fuel Storage,'' to increase fuel

assembly enrichment limits to 5.0 w/o U-235 while maintaining

Keff in the storage pools (spent fuel pool and new fuel

storage racks) less than 0.95. The May 15, 1997, supplement provided a

revised no significant hazards consideration determination that

superseded the licensee's determination noticed in the Federal Register

on April 23, 1997 (62 FR 19837).

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 this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a change to structures,

systems, or components that would affect the probability or

consequences of an accident previously evaluated in the PBNP Final

Safety Analysis Report (FSAR). The only relevant concern with

respect to increasing enrichment limits in the spent fuel pool and

new fuel storage racks is one of criticality. The proposed changes

use the same criticality limit used in the current Technical

Specifications. Therefore, margin to safe operation of Units 1 and 2

is maintained. The probability and consequences of an accident

previously evaluated are dependent on this criticality limit.

Because the limit will not change, the probability and consequences

of those accidents previously evaluated will not change.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes do not involve a change to the physical

structure of the spent fuel pool or of the plant. The proposed

increase in spent fuel pool and new fuel storage racks fuel assembly

enrichment limits maintains the margin to safe operation of Units 1

and 2 because the criticality limit for the spent fuel pool and new

fuel storage racks will not change. The enrichment increase does not

affect any of the parameters or conditions that contribute to the

initiation of any accidents. Because the criticality limit remains

the same, these changes have no effect on plant operation or on the

initiation of any accidents. Therefore, the proposed changes will

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

from any accident previously evaluated.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The proposed changes maintain the margin to safe operation of

Units 1 and 2. The margin of safety is based on the criticality

limit of the spent fuel pool and the new fuel storage racks. Because

this limit will not change, the margin of safety will not be

affected. Therefore, the proposed changes will not create a

significant reduction in a margin of safety.

[[Page 30648]]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

NRC Project Director: John N. Hannon

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: April 23, 1997

Description of amendment request: This request proposes to revise

Technical Specification 3/4.9.4, Containment Building Penetrations, and

its associated Bases section, to allow selected containment isolation

valves to be opened under administrative controls during periods of

core alterations or movement of irradiated fuel inside containment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change involves changes to the Technical

Specification requirements for containment closure which is an

accident mitigating feature. The changes would not affect the

likelihood of occurrence of any accidents previously evaluated. The

proposed change does not involve any hardware or plant design

changes. The containment leakage value is not assumed to be an

initiator of any analyzed event. Containment isolation valves and

temporary closure devices serve to limit the radiological

consequences of accidents. The proposed change would ensure the

service air and breathing air manual isolation valves will perform

their required containment closure function and will serve to limit

the consequences of a fuel handling accident as described in the

USAR, such that the results of the analyses in the USAR remain

bounding. In considering the consequences of a design basis fuel

handling accident inside containment, the assumptions in the

analysis take no credit for the containment as a barrier to prevent

the postulated release of radioactivity. For events that could occur

during CORE ALTERATIONS or movement of irradiated fuel assemblies,

containment closure is considered a defense-in-depth boundary to

prevent uncontrolled release of radioactivity. Additionally, the

proposed change does not impose any new safety analyses limits or

alter the plant's ability to detect and mitigate events. Therefore,

this change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or diffe

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