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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued, from June 19, 1999, through July 2, 1999. The

last biweekly notice was published on June 30, 1999 (64 FR 35199).

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 and

Directives Branch, Division of Administration 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 August 13, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

scheduled in the proceeding, but such an amended

[[Page 38023]]

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: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. 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.

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

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

Carolina

Date of amendment request: June 15, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to incorporate the performance-

based 10 CFR 50 Appendix J, Option B for Type A tests (containment

integrated leakage rate tests). Option B will be implemented for Type A

testing in accordance with NRC Regulatory Guide 1.163, ``Performance-

Based Containment Leak-Test Program,'' dated September 1995, and

Nuclear Energy Institute (NEI) Guideline 94-01, Revision 0, ``Industry

Guideline for Implementing Performance-Based Option of 10 CFR Part 50,

Appendix J,'' dated July 26, 1995. Type B and C testing (containment

penetration leakage tests) will continue to be performed in accordance

with 10 CFR 50 Appendix J, Option A.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The Harris Nuclear Plant (HNP) Type A testing history provides

justification for the proposed test schedule change to one test in a

10 year period. With the successful Type A tests of September 1992

and May 1997, and a greater than 24 month elapsed time between the

two tests, CP&L considers the requirement of two consecutive Type A

tests to have been met. This testing has affirmed the acceptable

reliability of the containment structure to minimize leakage as

designed, and provides assurance that its performance to

continuously function as designed is not challenged due to this test

schedule extension to once in 10 years.

This proposed change to revise the test schedule frequency does

not impact or alter the design of any system, structure or

component. The limit on allowable leakage is not increased. Type A

testing provides periodic verification of the leak tight integrity

of the containment and the components that penetrate the containment

structure. NUREG-1493, Section 10.1.2, ``Leakage-Testing

Intervals,'' states that reducing the frequency of Type A tests from

the current three per 10 years to one per 20 years was found to lead

to an imperceptible increase in risk.

Therefore, based on these considerations, and the previous

plant-specific Type A test results, the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes only incorporate the performance-based

testing approach authorized in 10 CFR 50 Appendix J, Option B, and

are justified based on previous plant-specific Type A test results.

Plant structures, systems, and components will not be operated in a

different manner as a result of these proposed changes and no

physical modifications to equipment are involved. The interval

extensions allowed by Option B of 10 CFR 50 Appendix J do not have

the potential for creating the possibility of a new or different

type of accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed changes do not change the allowable leak rate from

the containment; they only allow an extension of the interval

between the performance of Type A leak rate testing. NUREG-1493

provides the technical basis for the NRC's rulemaking to revise

containment leakage testing requirements for nuclear power reactors

in 10 CFR 50 Appendix J. NUREG-1493, Section 10.1.2, ``Leakage-

Testing Intervals,'' states that increasing the interval between

integrated leakage-rate tests is possible with minimal impact on

public risk.

Based on these considerations and the previous plant-specific

Type A test results, the proposed changes do 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

[[Page 38024]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

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

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

Date of amendment request: June 15, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification 4.7.D.6 by replacing the leakage limit

of 11.5 standard cubic feet per hour (scfh) for each main steam

isolation valve (MSIV) with a limit of 46 scfh on the total combined

leakage from all four main steam lines.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes to the Technical Specifications, Appendix

A, modifies the allowed MSIV leakage limit to an aggregate value

with no change to the total allowed leakage rate. This change does

not affect either the automatic or manual features that would close

the MSIVs. Performance of the leakage tests do not adversely affect

any accident previously evaluated. Consequently, this proposed

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The safety function of the MSIVs is to provide a timely steam

line isolation to mitigate the release of radioactive steam and

limit reactor inventory loss under certain accident and transient

conditions. The MSIVs are designed to automatically close whenever

plant conditions warrant main steam line isolation. Changing the

leakage limits to include an aggregate value does not affect the

isolation function. No new equipment will be installed or utilized,

and no new operating conditions will be initiated as a result of

this change. Therefore, the proposed change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

of safety?

The total allowed leakage rate for all MSIVs remains unchanged

at 46 scfh. Therefore, there will be no change in the types or

significant increase in the amounts of any effluents released

offsite, and, thus, the radiological analyses remain unchanged and

within the guidelines of 10 CFR 100 and General Design Criteria 19.

Therefore, these changes do not involve a significant reduction in

the margin of safety.

Therefore, based upon the above evaluation, ComEd has concluded

that these changes involve no significant hazards consideration. The

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

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

Date of amendment request: May 19, 1999.

Description of amendment request: The proposed amendments would

relocate Technical Specification Section \3/4\.4.4, ``Chemistry'' from

the TS to the Updated Final Safety Analysis Report and Administrative

Technical Requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes simplify the TS, meet regulatory

requirements for relocated TS, and implement the recommendations of

the NRC's Final Policy Statement on TS improvements. The Chemistry

requirements will be relocated to the Updated Final Safety Analysis

Report (UFSAR) and Administrative Technical Requirement that has

been incorporated into the UFSAR by reference. Future changes to

these requirements will be controlled by 10 CFR 50.59. The proposed

changes are administrative in nature and do not involve any

modification to any plant equipment or affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of any previously

evaluated accident.

Consequently, this proposed amendment does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety related system performs

its function. Therefore, this proposed TS amendment would not create

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

accident previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed amendment represents the relocation of current

requirements that are based on generic guidance or previously

approved provisions for other stations. The proposed changes are

administrative in nature and do not adversely affect existing plant

safety margins or the reliability of the equipment assumed to

operate in the safety analysis. The proposed changes have been

evaluated and found to be acceptable for use at Duane Arnold Energy

Center and Quad Cities Nuclear Power Station. Since the proposed

changes are administrative in nature, and are based on NRC accepted

provisions which have been adopted at other nuclear facilities, and

maintain the necessary levels of system reliability, the proposed

changes do not involve a significant reduction in the margin of

safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

Plant, Middlesex County, Connecticut

Date of amendment request: June 3, 1999.

Description of amendment request: The proposed amendment would

delete sections of the Technical Specifications that no longer apply to

the Haddam

[[Page 38025]]

Neck Plant's permanently shutdown and defueled condition; increase the

weight of loads allowed over the spent fuel pool; relocate certain

definitions and requirements from the Technical Specifications to the

Technical Requirements Manual (TRM), Connecticut Yankee Quality

Assurance Program (CYQAP), or the Radiological Effluent Monitoring and

Offsite Dose Calculation Manual (REMODCM); and correct typographical

errors, renumber sections, and repaginate the Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The licensee's summary of its analysis is presented

below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

In the present plant configuration, the reactor-related

accidents previously evaluated (i.e., LOCA [loss-of-coolant

accident], MSLB [main steamline break], etc.) are no longer

possible. The accidents previously evaluated that are still

applicable to the plant are fuel handling accidents and gaseous and

liquid radioactive releases. The following events are presently

considered as bounding of all other events:

--Fuel handling and cask drop accidents in the spent fuel building,

--Criticality in the spent fuel pool,

--Loss of spent fuel cooling,

--Resin fire (gaseous release), and

--Rupture of a tank containing radioactive liquid.

There is no significant increase in the probability of a fuel

handling accident since refueling operations have ceased, with a

corresponding decrease in the frequency of fuel movement. The

radiological consequences of a fuel handling accident, should one

occur, decrease the longer the spent fuel is allowed to decay. As

discussed previously, the spent fuel inventory of radioactive iodine

and noble gases [, with the exception of Kr-85,] have decayed more

than 20 half-lives since shutdown and are no longer a release

concern. With this reduced source the results of the fuel handling

accident show that the filters of Specification 3.9.12 are no longer

necessary. The allowed weight over the spent fuel pool is still less

than that previously [evaluated]. Therefore, there has been no

increase in the probability or consequences of a fuel handling or

cask drop accident.

Criticality controls are imposed by specifications \3/4\.9.13

and \3/4\.9.14 * * * [The requirements of these specifications have

not been changed.] Therefore, there has been no increase in the

probability or consequences of a criticality event.

Spent fuel cooling is maintained by keeping the pool temperature

below 150 deg.F. Should normal cooling be lost, the availability of

an abundant supply of water ensures that sufficient time is

available prior to boiling to restore cooling. This is controlled by

specifications \3/4\.9.11 and \3/4\.9.16 * * * [The requirements of

these specifications have not been changed. Technical specification

\3/4\.9.15 does not apply to the permanently defueled condition of

the plant. Therefore, there has been no increase in the probability

or consequences of a loss of cooling event.]

The probability of a gaseous or liquid radioactive release is

not changed by the proposed revisions. As the plant undergoes

decommissioning, the previous limiting events [such as a loss-of-

coolant-accident] are no longer applicable, and previous non-

limiting events [such as a resin fire] now become limiting. These *

* * events have not changed from how they might have occurred in the

past. The radiological consequences of a gaseous or liquid

radioactive release are bounded by the fuel handling accident during

defueled operation and a spent resin fire during processing of resin

from the reactor coolant system decontamination. The rupture of a

tank containing radioactive liquid was assessed and found to be

bounded by these events. With the plant defueled and permanently

shutdown, the demands on the radwaste systems are lessened since no

new radioisotopes are being generated by irradiation or fission.

Therefore, there is no increase in the probability or consequences

of a gaseous or liquid radioactive release.

* * * * *

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

from any accident previously evaluated.

The proposed changes are generally of an administrative nature

and do not have an effect on the physical plant. The events

considered bound other potential events and are considered the

limiting cases for potential gaseous or liquid releases to the

environment.

With the plant undergoing decommissioning, the types of

accidents one might be concerned with involve criticality of the

spent fuel, or draining of the spent fuel pool. None of the proposed

changes affect the possibility of such an event. Also, none of the

proposed changes could lead to a radiological release of a greater

magnitude than for the events considered, such as might occur with

the accumulation of a greater quantity of radioactive material in

one location, or with damage to a greater number of fuel assemblies

than considered in the fuel handling accident.

The proposed changes restrict the operations that can be

conducted at the plant, and do not permit any new type of activity

from what had previously been authorized. The effect on systems,

structures and components affected by the proposed changes have no

adverse impact on the storage of fuel nor on the processing of

radioactive wastes presently at the site. The present set of

limiting events are a subset of events previously considered.

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

different kind of accident from any accident previously considered.

* * * * *

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

The proposed changes have no impact on the analyses of

postulated design basis events remaining applicable to the Haddam

Neck Plant. Analysis of the limiting events show that their

consequences to the public are within the limits of 10 CFR [Part] 20

and the EPA PAGs [Environmental Protection Agency Protective Action

Guides]. The consequences to members of the operating staff are

within the limits of 10 CFR [Part] 50, Appendix A, General Design

Criterion 19, ``Control Room Habitability''. Therefore there is no

reduction in a margin of safety.

* * * * *

Based on the above, the proposed changes to the operating license

and technical specifications do not involve a reduction in the margin

of safety due to the reduced decay heat load, the decay of

radionuclides since shutdown, and by maintaining the heavy load

restriction.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut, 06457.

Attorney for licensee: Mr. J. A. Ritsher, Ropes & Gray, One

International Place, Boston, Massachusetts 02110-2624.

NRC Section Chief: Michael T. Masnik.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: June 10, 1999.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) to adopt a Ventilation Filter

Testing Program in TS Section 6.0, ``Administrative Controls,'' and

remove the specific ventilation filter surveillance requirements from

TS 3/4.6.4.4, ``Hydrogen Purge System,'' TS 3/4.6.5.1, ``Shield

Building Emergency Ventilation System,'' and TS 3/4.7.6.1, ``Control

Room Emergency Ventilation System.''

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

[[Page 38026]]

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 Number 1, in accordance with this change would:

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

accident previously evaluated because no change is being made to any

accident initiator. The replacement of the specific Technical

Specification (TS) ventilation filter testing Surveillance

Requirements for the Containment Hydrogen Purge System 3/4.6.4.4),

Shield Building Emergency Ventilation System (3/4.6.5.1), and the

Control Room Emergency Ventilation System (3/4.7.6.1), with a

reference to the newly created Ventilation Filter Testing Program

contained in TS Administrative Controls Section 6.8.4.f, Ventilation

Filter Testing Program, is a removal and relocation of certain TS

details. The proposed TS 6.8.4.f will, however, add controls to

maintain similar operation, maintenance, testing and system

operability for these three ventilation systems. The TS Bases

changes reflect the use of the Ventilation Filter Testing Program.

Therefore, it can be concluded that the proposed changes do not

involve a significant increase in the probability of an accident

previously evaluated.

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

accident previously evaluated because the proposed changes do not

affect accident conditions or assumptions used in evaluating the

radiological consequences of an accident. No physical alterations of

the DBNPS are involved, nor are plant operating methods being

changed. The proposed changes do not alter the source term,

containment isolation or allowable radiological releases.

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

accident from any accident previously evaluated because the proposed

changes do not change the way the plant is operated. No new or

different types of failures or accident initiators are being

introduced by the proposed changes.

3. Not involve a significant reduction in a margin of safety

because no inputs into the calculation of any Technical

Specification Safety Limit, Limiting Safety System Settings,

Technical Specification Limiting Condition for Operation, or other

previously defined margins for any structure, system, or component

important to safety are being affected by the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

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

NRC Section Chief: Anthony J. Mendiola.

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

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

Date of amendment request: May 10, 1999.

Description of amendment request: The proposed amendment would

correct the regulation referenced in Section 5.8, ``High Radiation

Area,'' of the Crystal River Unit 3 (CR-3) Improved Technical

Specifications (ITS). The ITS currently references 10 CFR 20, Paragraph

20.1601(2) and (3), whereas the correct reference is 10 CFR 20.1601(c).

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 change to the Crystal River Unit 3 (CR-3) Improved

Technical Specifications (ITS) is editorial in nature. The change

involves revising the incorrect reference in ITS Section 5.8.1 to

the correct Code of Federal Regulations reference that pertains to

controlling access to high radiation areas. The proposed ITS change

does not involve any change to plant design, operation, maintenance,

or procedures. As a result, no changes to the plant are being made

which would impact either the contributors to an accident or to the

consequences of an accident.

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

from any accident previously evaluated?

The proposed editorial change to the ITS does not involve any

changes to any plant structure, system, or component (SSC) or to its

operation or maintenance. There is no impact on any equipment that

would be considered as contributors to either new or different

accidents. Thus, the change to the ITS does not create the

possibility of a new or different kind of accident.

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

The proposed change to the ITS involves a reference change and

does not involve the design or operation of any plant SSC. No

changes to the methods for controlling access to high radiation

areas are proposed. No changes to the methods for controlling

personnel and/or activities in high radiation areas are proposed.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

this review, it appears that the three standards of 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 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC-A 5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Section Chief: Sheri R. Peterson.

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

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

Date of amendment request: May 17, 1999.

Description of amendment request: The proposed amendment would

revise a note in Surveillance Requirement (SR) 3.3.8.1 in the Crystal

River Unit 3 Improved Technical Specifications (ITS). The note

currently states that, when Emergency Diesel Generator (EDG) Loss Of

Power Start instrumentation is placed in an inoperable status solely

for performance of this surveillance, entry into associated Conditions

and Required Actions may be delayed for up to four hours provided the

two channels monitoring the Function for the bus are OPERABLE or

tripped. The proposed revision to the note states that entry into the

Conditions and Required Actions of ITS Section 3.3.8 is not required

provided the applicable Conditions and Required Actions of ITS Section

3.8.1, ``Electrical Power Systems, AC Sources--Operating,'' are entered

for the EDG being made inoperable. The proposed amendment would also

delete a superceded 60-day surveillance frequency and the associated

note which indicated that the frequency was not effective after

November 23, 1997.

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?

Proposed Improved Technical Specifications (ITS) Change A--Revision of

Surveillance Note

The note in Surveillance Requirement (SR) 3.3.8.1 involves the

timing for placing Crystal River Unit 3 (CR-3) into the applicable

Conditions and Required Actions when

[[Page 38027]]

performing the surveillance. The proposed revision will make the

note consistent with the actual method of performing the

surveillance at CR-3. The design and testing configuration does not

allow CR-3 to use the relief provided by the note. As a result, the

Conditions and Required Actions of ITS Section 3.3.8 are entered at

the start of the surveillance performance. The design and testing

configuration requires entry into ITS Section 3.8.1 Conditions prior

to performing SR 3.3.8.1. The revised note would require entering

the applicable Conditions and Required Actions of ITS Section 3.8.1

for one Emergency Diesel Generator (EDG) being inoperable. This

approach and the proposed note are conservative relative to the

current note. The proposed note results in no changes to the method

or to the timing of performing SR 3.3.8.1. Direct entry into ITS

Section 3.8.1 Conditions will achieve the same final ITS condition

as if Section 3.3.8 Conditions and Required Actions were entered.

Therefore, the probability of occurrence and the consequences of any

accident previously evaluated are unaffected by this change.

Proposed ITS Change B--Deletion of Frequency Note

The note under Frequency in SR 3.3.8.1 involves the period of

time that the 60 day surveillance frequency would be in effect. The

60 day frequency was a temporary extension that was needed to

implement modifications to the EDG during the 1997 CR-3 design

outage. This was a one-time extension of the frequency. The note

indicates this temporary nature of the 60 day frequency. Deleting

the note is an editorial change since the surveillance has reverted

back to its 31 day frequency and the note is no longer effective.

Because the proposed deletion of the note is an editorial change,

and no change is proposed to the current 31 day frequency, the

probability of occurrence and the consequences of any accident

previously evaluated are unaffected by this change.

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

from previously evaluated accidents?

Proposed ITS Change A--Revision of Surveillance Note

The proposed revision of the note in SR 3.3.8.1 involves only

the timing of entry into associated ITS Conditions and Required

Actions. No changes are proposed to the existing ITS Conditions and

Required Actions. The proposed change is conservative since it will

require entering the appropriate Conditions and Required Actions

immediately upon starting SR 3.3.8.1. Changing the timing for entry

into ITS Conditions and Required Actions does not create the

possibility of a new or different kind of accident from those

evaluated previously.

Proposed ITS Change B--Deletion of Frequency Note

Deletion of the note under SR 3.3.8.1 Frequency is an editorial

change since the note is no longer effective. The current frequency

for performing SR 3.3.8.1 is 31 days. This is the same frequency

that was in effect prior to the one-time, temporary change of the

frequency to 60 days. The editorial change of deleting the note that

is no longer effective does not create the possibility of a new or

different kind of accident from those evaluated previously.

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

Proposed ITS Change A--Revision of Surveillance Note

One manner in which a margin of safety related to a Surveillance

Requirement might be affected would be if entry into a Limiting

Condition for Operation (LCO) were delayed. The result of a delay in

entering an LCO would be an increase in the time before a Required

Action was taken, such as commencing a plant shutdown. Generally,

such allowed times reflected in ITS Required Actions are based on

some margin. Increasing the time allowed before starting a certain

Required Action might result in a reduction of a margin of safety.

However, the proposed ITS change allows entry into Section 3.8.1

Conditions immediately rather than after a delay. The proposed ITS

change does not change the final plant condition required by the

ITS. Therefore, the proposed ITS change does not result in a

reduction in a margin of safety.

Proposed ITS Change B--Deletion of Frequency Note

Another manner in which a margin of safety related to a

surveillance requirement might be affected would be if the frequency

of performance were changed. Generally, margin might be reduced if

the frequency were reduced (i.e., the interval between performing

surveillances were increased). This proposed editorial change to

delete the note in SR 3.3.8.1 Frequency does not result in a change

to the surveillance frequency. Thus, the proposed deletion of the

note does not affect the existing 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 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

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

33733-4042.

NRC Section Chief: Sheri R. Peterson.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: June 11, 1999.

Description of amendment request: The proposed amendment makes

various plant organization title changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

proposed changes are administrative in nature and do not affect

assumptions contained in plant safety analyses, the physical design

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

Specifications that preserve safety analysis assumptions. None of

the proposed changes involve a physical modification to the plant, a

new mode of operation or a change to the UFSAR [Updated Final Safety

Analysis Report] transient analyses. No Technical Specification

Limiting Condition for Operation, Action statement or Surveillance

Requirement is affected by any of the proposed changes. These

proposed changes do not reduce the level of qualification, authority

or accountability associated with the affected Technical

Specification responsibilities. Further, the proposed changes do not

alter the design, function, or operation of any plant component.

Therefore, the proposed amendment does not affect the probability of

occurrence or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. The proposed changes

are administrative in nature and do not affect assumptions contained

in plant safety analyses, the physical design and/or modes of plant

operation defined in the plant operating license, or Technical

Specifications that preserve safety analysis assumptions. The

proposed changes do not introduce a new mode of plant operation or

surveillance requirement, nor involve a physical modification to the

plant. The proposed changes do not alter the design, function, or

operation of any plant components. Therefore, the proposed amendment

does not affect the possibility of a new or different kind of

accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed

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

proposed changes are administrative in nature. There is no reduction

in the organization position qualifications, authority and

accountability associated with the affected Technical Specification

responsibilities. None of the proposed changes involve a physical

modification to the plant, a new mode of operation or a change to

the UFSAR transient analyses. No Technical Specification Limiting

Condition for Operation, Action statement, or Surveillance

Requirement is affected. Therefore, the proposed amendment does not

reduce the margin of safety.

[[Page 38028]]

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 Section Chief: S. Singh Bajwa.

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

County, Iowa

Date of amendment request: April 12, 1999.

Description of amendment request: The proposed amendment would

revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)

Surveillance Requirement (SR) 3.6.1.3.7 to allow a representative

sample of reactor instrumentation line excess flow control valves

(EFCV) to be tested every 24 months, instead of testing each EFCV every

24 months.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The current SR frequency requires each reactor instrumentation

line EFCV to be tested every 24 months. The EFCVs at DAEC are

designed so that they will not close accidentally during normal

operation, will close if a rupture of the instrument line is

indicated downstream of the valve, can be reopened when appropriate,

and have their status indicated in the control room (reference DAEC

UFSAR [updated final safety analysis report] 1.8.11). This proposed

change allows a reduced number of EFCVs to be tested every 24

months. There are no physical plant modifications associated with

this change. Industry operating experience demonstrates a high

reliability of these valves. Neither EFCVs nor their failures are

capable of initiating previously evaluated accidents; therefore

there can be no increase in the probability of occurrence of an

accident regarding this proposed change.

Instrument lines connecting to the Reactor Coolant Pressure

Boundary (RCPB) with EFCVs installed also have a flow-restricting

orifice upstream of the EFCV. The consequences of an unisolable

rupture of such an instrument line [have] been previously evaluated

in response to Regulatory Guide (RG) 1.11 (DAEC UFSAR 1.8.1.1). That

evaluation assumed a continuous discharge of reactor water for the

duration of the detection and cooldown sequence (3.5 hours).

Therefore, although not expected to occur as a result of this

change, the postulated failure of an EFCV to isolate as a result of

reduced testing is bounded by this previous evaluation. Therefore,

there is no increase in the previously evaluated consequences of the

rupture of an instrument line and there is no potential increase in

the consequences of an accident previously evaluated as a result of

this change.

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

new or different kind of accident from any accident previously

evaluated.

This proposed change allows a reduced number of EFCVs to be

tested each operating cycle. No other changes in requirements are

being proposed. Industry operating experience demonstrates the high

reliability of these valves. The potential failure of an EFCV to

isolate by the proposed reduction in test frequency is bounded by

the previous evaluation of an instrument line rupture. This change

will not physically alter the plant (no new or different type of

equipment will be installed). This change will not alter the

operation of process variables, structures, systems, or components

as described in the safety analysis. Thus, a new or different kind

of accident will not be created.

3. The proposed amendment will not involve a significant

reduction in a margin of safety.

The consequences of an unisolable rupture of an instrument line

[have] been previously evaluated in response to RG 1.11 (reference

DAEC UFSAR 1.8.1.1). That evaluation assumed a continuous discharge

of reactor water for the duration of the detection and cooldown

sequence (3.5 hours). The only margin of safety applicable to this

proposed change is considered to be that implied by this evaluation.

Since a continuous discharge was assumed in this evaluation, any

potential failure of an EFCV to isolate postulated by this reduced

testing frequency is bounded and 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: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

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

NRC Section Chief: Claudia M. Craig

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

County, Iowa

Date of amendment request: April 30, 1999.

Description of amendment request: The proposed amendment would

revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)

Surveillance Requirement (SR) 3.4.3.1 to revise the safety function

lift setpoint tolerance limits for the main safety valves (SVs) and the

safety/relief valves (SRVs). The current tolerance bands for the SVs

and SRVs would be revised from -3% to +1% to a new band of plus or

minus 3%.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change allows an increase in the as-found SV and

SRVs safety mode setpoint tolerance, determined by test after the

valves have been removed from service, from +1%/-3% to plus or minus

3%.

The proposed change does not alter the TS requirements on the

nominal SV or SRV safety mode lift setpoints, the SRV relief mode

setpoints, the required frequency for the SV or SRV lift setpoint

tests, or the number of SVs and SRVs required to be operable.

Consistent with current requirements, this change continues to

require that these valves be adjusted to within plus or minus 1% of

their nominal lift setpoints following testing. This change does not

change the behavior and operation of any SV or SRV and therefore has

no significant impact on reactor operation. It also has no

significant impact on response to any perturbation of reactor

operation including transients and accidents previously analyzed in

the Updated Final Safety Analysis Report (UFSAR).

This change does not involve physical changes to the valves, nor

does it change the operating characteristics or safety function of

the valves. The proposed TS revision involves no significant changes

to the operation of any systems or components in normal or accident

operating conditions and no changes to existing structures, systems,

or components. Therefore these changes will not increase the

probability of an accident previously evaluated.

Generic considerations related to the change in setpoint

tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure

Relief Technical Specification Revision Licensing Topical Report,''

and were reviewed and approved by the NRC in a Safety Evaluation

(SE) dated March 8, 1993. The plant specific evaluations, required

by the NRC's SE and performed to support this proposed change, show

that there is adequate

[[Page 38029]]

margin to the design core thermal limits and to the reactor vessel

pressure limits using a plus or minus 3% setpoint tolerance. They

also show that operation of the high pressure coolant injection

(HPCI) and reactor core isolation cooling (RCIC) systems will not be

adversely affected and the containment response from a loss of

coolant accident will be acceptable. The plant systems associated

with these proposed changes will still be capable of meeting all

applicable design basis requirements and retain the capability to

mitigate the consequences of accidents described in the UFSAR.

Therefore, these changes will not involve a significant increase in

the consequences of any accident previously evaluated.

Therefore, the proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

These proposed changes were developed in accordance with the

provisions contained in the NRC SE, dated March 8, 1993, for the

``BWR Owners Group Inservice Pressure Relief Technical Specification

Revision Licensing Topical Report,'' NEDC-31753P. The revised SV and

SRV setpoint tolerance limit will not adversely impact the operation

of any safety related component or equipment. Since the proposed

changes involve no significant hardware changes, no significant

changes to the operation of any systems or components, and no

changes to existing structures, systems, or components, there can be

no impact on the occurrence of any accident.

The proposed change would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change to allow an increase in the SV and SRV safety

mode setpoint tolerance from +1%/-3% to plus or minus 3% does not

alter the nominal SV or SRV lift setpoints or the number of SVs or

SRVs required to be operable. This change does not involve physical

changes to the valves, nor does it change the operating

characteristics or the safety function of the valves. The proposed

change does not involve a physical alteration of the plant. No new

or different equipment is being installed. There is no alteration to

the parameters within which the plant is normally operated. As a

result no new failure modes are being introduced. There are no

changes in the methods governing normal plant operation, nor are the

methods utilized to respond to plant transients altered.

Therefore, the proposed amendment will not create the

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

previously evaluated.

3. The proposed amendment will not involve a significant

reduction in a margin of safety.

The proposed change does not involve a significant reduction in

a margin of safety. Establishment of the plus or minus 3% SV and SRV

setpoint tolerance limit will not adversely impact the operation of

any safety related component or equipment. Engineering evaluations

concluded that there are no significant impacts on fuel thermal

limits, safety related systems, structures or components, and no

significant impact on the accident analyses associated with the

proposed changes.

The margin of safety is established through the design of the

plant structures, systems, and components, the parameters within

which the plant is operated, and the establishment of the setpoints

for the actuation of equipment relied upon to respond to an event.

The proposed change does not significantly impact the condition or

performance of structures, systems, and components relied upon for

accident mitigation. The proposed change does not significantly

impact any safety analysis assumptions or results.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

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

NRC Section Chief: Claudia M. Craig.

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

County, Iowa

Date of amendment request: May 10, 1999.

Description of amendment request: The proposed amendment would

revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)

Section 2.1.1.2, to revise the Safety Limit Minimum Critical Power

Ratio (SLMCPR) to support operation with GE-12 fuel with a 10x10 pin

array.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

There is no change to any plant equipment other than the fuel.

The SLMCPR protects the fuel in accordance with the design bases.

The SLMCPR calculations limit the bundle power to ensure the

critical power ratio remains unchanged. Therefore, there is not an

increase in the probability of transition boiling. The basis of the

SLMCPR calculation remains the same, ensuring that greater than

99.9% of all fuel rods in the core avoid transition boiling if the

limit is not violated. Therefore, there is no increase in the

probability of occurrence of a previously evaluated accident.

The fundamental sequences of accidents have not been altered.

The Minimum Critical Power Ratio (MCPR) Operating Limits are

selected such that potentially limiting accidents do not cause the

MCPR to decrease below the SLMCPR anytime during the accident.

Therefore, there is no impact on any of the limiting accidents.

Therefore there is no increase in the consequences of any accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The SLMCPR values are designed to ensure that fuel damage from

transition boiling does not occur in at least 99.9% of the fuel rods

as a result of the limiting postulated accident. The values are

calculated in accordance with NRC-approved General Electric methods.

The approved General Electric methods are comprehensive for ensuring

that fuel designs will perform within acceptable bounds. The SLMCPR

ensures that the fuel is protected in accordance with the design

basis. The function, location, operation, and handling of the fuel

remain unchanged. Therefore, the possibility of a new or different

kind of accident is not created.

3. The proposed amendment will not involve a significant

reduction in a margin of safety.

The SLMCPR values do not alter the design or function of any

plant system. The new values were calculated using NRC-approved

methods to maintain the same margin of safety as presently exists

for the prevention of transition boiling. At least 99.9% of the fuel

rods will avoid transition boiling if the SLMCPR is not violated.

Therefore, a significant reduction in a margin of safety is not

involved.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

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

NRC Section Chief: Claudia M. Craig.

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

County, Iowa

Date of amendment request: May 10, 1999.

Description of amendment request: The proposed amendment would

revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)

to: (1)

[[Page 38030]]

insert NOTE for Limiting Condition for Operation (LCO) 3.7.4 that would

allow intermittent opening of the control building boundary under

administrative control; (2) add a CONDITION, REQUIRED ACTION and

COMPLETION TIME to LCO 3.7.4 for when both standby filter unit (SFU)

subsystems are inoperable due to inoperable control building boundary

in MODES 1, 2, and 3; (3) re-letter items in LCO 3.7.4 for consistency;

and (4) revise LCO 3.7.4 CONDITION D (new CONDITION E) to add ``for

reasons other than CONDITION B.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Requiring the plant to enter LCO 3.0.3 when the control building

pressure envelope is not intact is excessively restrictive. This

change provides less restrictive requirements for operation of the

facility. These less restrictive requirements do not result in

operation that will increase the probability of initiating an

analyzed event. The proposed change is acceptable because of the low

probability (less than 3.04 x 10-8) of a DBA [design

basis accident] occurring during the 24 hour Completion Time, and

the availability the SFU system to provide a filtered environment

(albeit with potential control room in-leakage).

Intermittent opening of the control building boundary requires

controls which consist of stationing a dedicated individual at the

opening who is in continuous communication with the control room.

This individual will have a method to rapidly close the opening when

a need for control room isolation is indicated. For entry and exit

through doors the administrative control is performed by the person

entering or exiting the area. As a result, the consequences of any

accident previously evaluated are not significantly increased.

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

new or different kind of accident from any accident previously

evaluated. This change does not involve new or different equipment

being installed at the facility. The proposed change is acceptable

because of the low probability (less than 3.04 x 10-8) of

a DBA occurring during the 24 hour Completion Time, and the

availability of the SFU system to provide a filtered environment

(albeit with potential control room in-leakage).

Intermittent opening of the control building boundary requires

controls which consist of stationing a dedicated individual at the

opening who is in continuous communication with the control room.

This individual will have a method to rapidly close the opening when

a need for control building isolation is indicated. For entry and

exit through doors the administrative control is performed by the

person entering or exiting the area.

3. The proposed amendment will not involve a significant

reduction in a margin of safety. Requiring the plant to enter LCO

3.0.3 when the control room ventilation envelope is not intact is

excessively restrictive. The proposed change is acceptable because

of the low probability (less than 3.04 x I0-8) of a DBA

occurring during the 24 hour Completion Time.

Intermittent opening of the control room boundary requires

controls which consist of stationing a dedicated individual at the

opening who is in continuous communication with the control room.

This individual will have a method to rapidly close the opening when

a need for control building isolation is indicated. For entry and

exit through doors the administrative control is performed by the

person entering or exiting the area.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

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

NRC Section Chief: Claudia M. Craig.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: June 8, 1999.

Description of amendment request: The proposed change corrects the

described method by which the Standby Gas Treatment system heaters are

to be tested. This change is necessary because the reference provided

in Technical Specification Section 5.5.7e is in error.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The correction of an error and clarification of a testing

method does not alter any of the precursors assumed in the CNS

[Cooper Nuclear Station] accident analysis. The proposed wording for

testing SGT [Standby Gas Treatment] heaters is in accordance with

ASME N510-1989, Section 14.5.1, ``Testing of Nuclear Air Treatment

Systems.'' Since the proposed change does not affect this portion of

plant design and operation, the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change will not create the possibility of a new or

different kind of accident than evaluated in the Updated Safety

Analysis Report (USAR). The proposed change does not result in any

physical change to CNS structures, systems, or components, nor does

it change the fit, form, or function of any equipment or components

taken credit for in the accident analyses described in the USAR.

Therefore, correction of a test reference and specific description

of the testing method for the SGT heaters does not create the

possibility of a new or different kind of accident.

The proposed change will not create a significant reduction in

the margin of safety. The proposed change does not alter the design

or administrative controls necessary to ensure the required

performance of the physical barriers during acticipated operational

occurrences and postulated accidents. This conclusion is based on

the fact that the proposed change corrects an erroneous reference,

conforms to industry standards, and is consistent with past and

current operating practice at CNS; therefore, the proposed change

does not create 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: June 15, 1999.

Description of amendment request: The proposed change would allow

the use of the service water (SW) system to directly supply cooling

water to the reactor equipment cooling (REC) system during a loss-of-

coolant accident (LOCA) event. The present maximum allowable REC water

leakage rate is based on the requirement that there will be sufficient

water in the REC surge tank to allow the REC system to fulfill its

safety function for 30 days post-LOCA condition. A proposed Updated

Safety Analysis Report (USAR) revision would allow Cooper Nuclear

Station (CNS) to

[[Page 38031]]

revise the maximum allowable REC system leakage during normal power

operation such that the REC system surge tank would assure that the REC

would fulfill its safety function for at least the first 7 days

following a large break LOCA. The SW system would fulfill the safety

functions of the REC system, if required, for the remaining duration of

the accident.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below. The licensee states that the

proposed request:

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

consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in

the probability of an accident previously evaluated in the USAR

since there are no hardware changes associated with this USAR

change. Procedure changes associated with this USAR change are

limited to direction on which division of SW/REC backup to initiate

first, and incorporation of new system leakage limits into

surveillance procedures.

The proposed change also does not involve a significant increase

in the consequences of an accident previously evaluated in the USAR.

This conclusion is based on the safety evaluation (Attachment 2 [of

the June 15, 1999, application]) which demonstrates that the SW

system will fulfill the safety functions of the REC system in a post

LOCA condition and thus the proposed change will not affect the

performance and reliability of the REC system. The emergency systems

cooled by the REC system, the ECCS [emergency core cooling] systems

and their room coolers, will therefore also fulfill their safety

function when directly supplied by the SW system.

2. Does not create the possibility for a new or different kind

of accident from any accident previously evaluated.

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

different kind of accident from any accident previously evaluated in

the USAR. The proposed license amendment does not introduce any new

equipment or hardware changes. It does, however, allow the SW system

to perform a different type of function than it is presently

licensed to perform in a post LOCA condition. This SW system post

LOCA function has been previously demonstrated to fulfill the

functions of the REC in a non LOCA emergency shutdown which are the

same as the functions required following a LOCA.

3. Does not create a significant reduction in the margin of

safety.

The proposed activity does not involve a significant reduction

in the margin to safety. The safety evaluation (Attachment 2)

demonstrates that the SW system will perform the required REC post

LOCA functions. There is an added required operator action which is

to align the SW system to directly supply cooling water to the REC

critical loops. As discussed in the safety evaluation [of the June

15, 1999, application], this action can be performed from the main

control room utilizing one control switch and there is sufficient

control room indication for the operator to be alerted to the need

for the use of service water backup. There is also sufficient time

for the operator to perform the task. Trending (prior to a

postulated LOCA) routinely provides the control room operator with

REC system leakage information. In a post LOCA situation, this

leakage information would assist the operator in taking timely

action to initiate the service water back-up before the need is

alarmed in the control room.

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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: June 15, 1999.

Description of amendment request: The purpose of the requested

license amendment is to revise the Updated Safety Analysis Report

(USAR) to incorporate the latest analysis to demonstrate adequate net

positive suction head (NPSH) for the low pressure emergency core

cooling system (ECCS) pumps following a large break loss-of-coolant

accident (LOCA). Specifically, the change would allow (1) reliance on a

slightly larger amount of containment overpressure for residual heat

removal (RHR) and core spray (CS) pump operation during worst-case

long-term LOCA conditions (greater than 1000 seconds) while still

maintaining original license margins of 3 and 6 pounds per square inch

(psi), respectively, for the difference between minimum available

containment pressure and the pressure required for minimum pump NPSH,

(2) reliance on a small amount of containment overpressure for CS pump

runout during worst-case short-term LOCA conditions (less than 10

minutes) while still maintaining an adequate pressure margin of at

least 5 psi, and (3) the use of ANS 5.1 decay heat model in the USAR

Section 5.2.6 as currently presented based on analysis justifying the

use of this model as described in the amendment request.

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. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change does not involve an increase in the

probability of an accident previously evaluated in the USAR. There

are no changes being proposed to the maintenance, operation, or

design of plant systems or equipment postulated to initiate

accidents or transients.

The proposed change does not involve an increase in the

consequences of an accident previously evaluated in the USAR. This

conclusion is based on the conclusions of the safety evaluation

(Attachment 2 [of the June 15, 1999, application]). This safety

evaluation demonstrates that the containment overpressure is

sufficiently conservative, and that the calculated margins between

the available containment overpressure and the overpressure required

to assure adequate low pressure ECCS pump NPSH are such that ECCS

pump operation, as credited in the CNS [Cooper Nuclear Station]

accident analysis, remains unchanged.

2. Does not create the possibility for a new or different kind

of accident from any accident previously evaluated.

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

different kind of accident from any accident previously evaluated in

the USAR. The proposed license amendment does not introduce any new

equipment or hardware changes. The attached safety evaluation

demonstrates that the only equipment affected by this License

Amendment are the low pressure ECCS pumps and that these will retain

their ability to function following a LOCA.

3. Does not create a significant reduction in the margin of

safety.

The proposed activity does not involve a significant reduction

in the margin of safety. The safety evaluation (Attachment 2)

demonstrates that, although there is an increased reliance on

containment overpressure to assure adequate low pressure ECCS pump

NPSH, there remains sufficient margin to provide confidence that the

ECCS pumps will operate as required. Sufficient margin is

demonstrated with the added conservatism of a 2-sigma (2 standard

deviation) uncertainty in the decay heat model, increased suction

strainer debris loading, increased RHR heat exchanger tube plugging

margin, and increases in SW [Service Water] and Suppression Pool

temperatures. The minimum margin available between available

overpressure and required overpressure is at least 5 psi for CS

(just prior to 10 minutes) and at least 3 psi for RHR (well after 10

minutes).

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

this

[[Page 38032]]

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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Robert A. Gramm.

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

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

Date of amendment request: June 23, 1999.

Description of amendment request: The proposed change to the

Technical Specifications would increase the allowed outage time for the

Control Room Air Conditioning Subsystem from 30 days to 60 days, on a

one-time basis only, for each train.

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 operational requirements for the Control Room Air

Conditioning Subsystems (CRACS) are contained in Technical

Specification 3.7.6.2 ``Control Room Subsystems Air Conditioning.''

This Limiting Condition for Operation (LCO) requires that two

independent Control Room Air Conditioning Subsystems (trains) be

operable during all modes of operation. The LCO action statement for

operational modes 1, 2, 3 and 4, with one Control Room Air

Conditioning Subsystem inoperable, states: ``restore the inoperable

system to operable status within 30 days or be in at least Hot

Standby [Mode 3] within the next 6 hours and in Cold Shutdown within

the following 30 hours.'' The LCO action statement for operational

modes 5 and 6 with one Control Room Air Conditioning Subsystem

inoperable, states: ``restore the inoperable system to operable

within 30 days or initiate and maintain operation of the remaining

OPERABLE Control Room Air Conditioning Subsystem or immediately

suspend all operations involving CORE ALTERATIONS or positive

reactivity changes.''

The proposed change adds the following note: ``* For cycle 7,

the allowable outage time may be extended to 60 days, on a one-time

basis, for each train, to implement modifications to the control

room air conditioning subsystems. The provisions of specifications

3.0.4 and 4.0.4 are not applicable during the implementation of

modifications to the air conditioning subsystems.''

This change is a one-time only change to Technical Specification

3.7.6.2 in order to facilitate the installation of a design change

to the CRACS during the present operating cycle. This change will

not affect the existing 30 [day] AOT period presently in place in

Technical Specification 3.7.6.2 which requires specific actions in

the event that the CRACS is determined to be inoperable for any

other reason. The design basis accidents are not affected as a

result of the proposed one-time change to the Technical

Specifications. The CRACS are support subsystems which can only

contribute to the initiation of an accident if the whole function is

lost. The plant would be required to shutdown before this occurred.

The proposed change does not adversely affect accident initiators or

precursors nor alter the design assumptions, conditions,

configuration of the facility (other than the CRACS) or the manner

in which the plant is operated nor does it adversely affect the

response of the plant to a transient or accident. This one-time

change is to be utilized only during the present operating cycle

(cycle 7) in order to facilitate the implementation of a design

change to modify the existing safety-related refrigerant subsystems

(one train at a time) and replace them with safety-related chilled

water subsystems. This design change is being implemented to improve

the overall reliability of the safety-related subsystems.

The consequences of an extended loss of the operating CRACS and

the non-safety related chilled water subsystem, during all modes of

operation, would result in a slow gradual rise in control room

temperature. The temperature of the control room is normally

maintained between 70 to 72 deg.F at the discretion of the Unit

Shift Supervisor utilizing a non-safety-related train of CRACS. In

the event that the control room temperature increased to a

temperature greater than 75 deg.F, plant procedures require starting

other equipment in the non-safety-related chilled water subsystem or

a safety-related train of CRACS to restore control room temperature

to its normal operating band. In the unlikely event that the non-

safety-related chilled water subsystems and the operable safety-

related train of CRACS fail during the proposed 60 day AOT period,

Technical Specification 3.7.6.2 would require that actions be

commenced to place the plant in a shutdown condition. Additionally,

alternative actions to reduce control room temperature could also be

initiated as identified in a plant procedure. It has been

conservatively determined that safety-related equipment in the

control room can be operated continuously up to 90\F in an

environment without affecting the capability of the equipment.

The exception to specifications 3.0.4 and 4.0.4 as stated in the

proposed one-time change to Technical Specification 3.7.6.2 will not

involve an increase in the probability or consequences of an

accident. TS 3.0.4 prohibits entry into a mode when the conditions

for the LCO are not met and the associated action(s) requires a

shutdown if they are not met within a specified time interval.

Surveillance Requirement 4.0.4 prohibits entry into a mode unless

the associated surveillance requirement(s) has been performed within

the stated interval. During the implementation of the modification,

when one safety-related train of CRACS is inoperable, it is possible

that a plant shutdown could occur due to reasons unrelated to the

planned modifications of the CRACS. As stated above, the CRACS are

support subsystems which do not contribute to the initiation of any

accident previously evaluated. Entry of the plant into an

operational mode from a shutdown mode as a result of the proposed

modification does not adversely affect accident initiators or

precursors nor alter the design assumptions, conditions,

configuration of the facility (other than the CRACS) or the manner

in which the plant is operated nor does it adversely affect the

response of the plant to a transient or accident. The functions of

the CRACS to provide a controlled environment inside of the control

room complex to ensure the comfort of the plant operators and to

ensure adequate climate conditions for the operability of equipment

will not be impaired in any way as a result of a plant mode change.

The remaining actions identified in TS 3.7.6.2 are unchanged as a

result of the proposed change. The risk significance involved with

removing a safety-related train of the CRACS during power operation

or during refueling conditions is low based on the short period (60

days per train) and consequences of losing this function. The CRACS

is excluded from modeling in the Seabrook Station Probabilistic Risk

Assessment (PRA) due to its extremely low risk significance.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

from any accident previously evaluated.

The proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated

since it is a support system and the loss of function will require a

plant shutdown. The proposed change adds the following note which

pertains to both affected action statements: ``* For cycle 7, the

Allowable Outage Time may be extended to 60 days, on a one-time

basis, for each train during the implementation of modifications to

the control room air conditioning subsystems. The provisions of

specifications 3.0.4 and 4.0.4 are not applicable during the

implementation of modifications to the air conditioning

subsystems.'' As previously identified, this change is a one-time

only change to Technical Specification 3.7.6.2 in order to

facilitate the installation of a design change to CRACS during the

present operating cycle.

The CRACS are support subsystems which do not contribute to the

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

evaluated nor is it used to mitigate the consequences of a transient

or accident. The functions of the CRACS are to provide a controlled

environment inside of

[[Page 38033]]

the control room complex to ensure the comfort of the plant

operators and to ensure adequate climate conditions for the

operability of equipment. The CRACS consists of two independent

safety-related air conditioning trains that provide cooling of

recirculated control room air. Due to previous reliability problems

with the CRACS, an additional non-safety chilled water subsystem has

been installed to provide control room cooling on a continuous

basis. Baseload operation of the non-safety related chilled water

subsystem to provide control room cooling reduces the operational

load on the safety-related refrigerant trains.

Implementation of the modification to the CRACS subsystems

during the 60 day AOT duration in no way affects the availability of

the non-safety-related chilled water subsystem or the operable

safety-related train of the CRACS to meet the control room cooling

requirements. The proposed modification removes freon from the

control room complex and the quantity of chilled water in the closed

loop system is too small to become a flood hazard. The consequences

of an extended loss of the operating CRACS train and the non-safety

related chilled water subsystem would result in a slow gradual rise

in control room temperature. In the event that control room

temperature increased to a temperature greater than 75 deg.F, plant

procedures require starting either the non-safety-related chilled

water subsystem or a safety-related train of CRACS to restore

control room temperature. Additionally, in the unlikely event of a

loss of the non-safety related chilled water subsystem and the

operable safety-related train of the CRACS, Technical Specification

3.7.6.2 would require that actions be taken to place the plant in a

shutdown condition.

It has been conservatively determined that safety-related

equipment in the control room can be operated continuously in an

environment up to 90 deg.F without affecting the capability of the

equipment. This proposed change will not affect the existing 30 day

AOT period presently in place in Technical Specification 3.7.6.2

which requires specific actions in the event that the CRACS is

determined to be inoperable for any other reason.

The exception to specifications 3.0.4 and 4.0.4 as stated in the

proposed one-time change to Technical Specification 3.7.6.2 will not

involve the creation of an accident of any type. During the

implementation of the proposed modification, when one safety-related

train of CRACS is inoperable, it is possible that a plant shutdown

could occur due to reasons unrelated to the planned modifications of

the CRACS. Entry of the plant into an operational mode from a

shutdown mode as a result of the proposed modification does not

adversely affect accident initiators or precursors nor alter the

design assumptions, conditions, configuration of the facility or the

manner in which the plant is operated nor the manner that it

responds to a transient or accident. The functions of the CRACS to

provide a controlled environment inside of the control room complex

to ensure the comfort of the plant operators and to ensure adequate

climate conditions for the operability of equipment will not be

impaired in any way as a result of a plant mode change. The

remaining actions identified in TS 3.7.6.2 are unchanged as a result

of the proposed change.

Therefore, the proposed change 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.

The proposed one-time change to Technical Specification 3.7.6.2

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

The functions of the CRACS are to provide a controlled environment

inside of the control room complex to ensure the comfort of the

plant operators and to ensure adequate climate conditions for the

operability of equipment. The CRACS consists of two independent

safety-related air conditioning trains that provide cooling of

recirculated control room air. Additionally, the Seabrook Station

design incorporates the use of a non-safety chilled water subsystem

(which is not within the scope of the Technical Specifications) to

provide baseload cooling of the control room on a continuous basis.

Implementation of the modification to the CRACS subsystems

during the 60 day AOT duration does not result in a significant

reduction in the plant margin of safety. As previously identified,

the CRACS is a support subsystem and the existing Technical

Specifications will require a plant shutdown on a loss of function.

The risk significance involved with removing a safety-related train

of the CRACS is extremely low based on the short period (60 days per

train) and the consequences of losing this function. The potential

that the non-safety-related chilled water subsystem and the operable

safety-related train of CRACS simultaneously fail during the

proposed 60 day AOT period of each safety-related train (120 days

total) is considered unlikely. In the event that control room

temperature increased to a temperature greater than 75 deg.F, plant

procedures require starting either the non-safety-related chilled

water subsystem or a safety-related train of CRACS to restore

control room temperature. Additionally, in the unlikely event of a

loss of the non-safety related subsystem and the operable safety-

related train of the CRACS, Technical Specification 3.7.6.2 would

require that actions be taken to place the plant in a shutdown

condition. Alternative actions to reduce control room temperature

could also be initiated as identified in a plant procedure. It has

been conservatively determined that safety-related equipment in the

control room can be operated continuously in an environment up to

90 deg.F without affecting the capability of the equipment.

The exception to specifications 3.0.4 and 4.0.4 as stated in the

proposed one-time change to Technical Specification 3.7.6.2 will not

reduce the margin of safety. During the implementation of the

proposed modification, when one safety-related train of CRACS is

inoperable, it is possible that a plant shutdown could occur due to

reasons unrelated to the planned modifications of the CRACS. Entry

of the plant into an operational mode from a shutdown mode as a

result of the proposed modification does not adversely affect

accident initiators or precursors nor alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. The functions of the CRACS to provide a

controlled environment inside of the control room complex to ensure

the comfort of the plant operators and to ensure adequate climate

conditions for the operability of equipment will not be impaired in

any way as a result of a plant mode change. The remaining actions

identified in TS 3.7.6.2 are unchanged as a result of the proposed

change.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

CT 06141-0270.

NRC Section Chief: James W. Clifford.

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

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

Date of amendment request: May 17, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) section 4.4.6.2.2.e to replace the

reference to American Society of Mechanical Engineers (ASME) Code

paragraph IWV-3472(b) which pertains to the frequency of leakage rate

testing for 6-inch, nominal pipe size valves and larger with the

requirement that the surveillance interval and frequency of

surveillance leakage rate testing for these valves be performed

pursuant to the requirements of TS 4.0.5, ``Operations and Surveillance

Requirements.''

Basis for proposed no significant hazards consideration

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

provided the NRC its analysis of the issue of no significant hazards

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

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

presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Eliminating the reference to ASME Code paragraph IWV-3427(b) and

[[Page 38034]]

performing pressure isolation valve (PIV) testing pursuant to TS 4.0.5

does not change the test conditions for PIV leakage testing and is

consistent with the currently analyzed configurations. This change

eliminates an unnecessary test requirement and incorporates

Westinghouse Owner's Group (WOG) Standard Technical Specifications

(STS) frequency requirements that are deemed to substantially reduce

the probability of an intersystem loss-of-coolant-accident. This change

in testing frequency requirements does not affect the accident

mitigation capabilities of the reactor coolant system (RCS) PIVs. This

change is bounded by existing accident analyses. Therefore, it is

concluded that, with the reduced probability of previously analyzed

accidents, and no effect on accident mitigation, the proposed revision

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any accident previously evaluated.

Eliminating the IWV-3427(b) trending for 6-inch and larger valves

(and the accompanying increased frequency testing requirement) does not

significantly change actual testing frequencies since the frequencies

continue to be addressed by the remaining TS requirements. This change

does not affect the ability of a PIV to perform its required RCS

pressure isolation safety function of limiting RCS leakage to prevent

overpressure failure of attached low pressure systems. The frequency of

testing or the testing itself are not initiating events to postulated

accidents. Therefore, the proposed revision does not create the

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

evaluated.

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

There is no impact on the Margin of Safety as defined in the bases

of any TS, the Updated Final Safety Analysis Report, or other licensing

basis commitments resulting from the elimination of the reference to

ASME Code paragraph IWV-3427(b). Periodic surveillances provide

continued assurance in the capability of safety related equipment to

perform its design safety (accident mitigating) function and are not

used to establish the margin of safety for accident mitigation.

Therefore, the frequency of surveillance testing of the PIVs has no

impact on the margins of safety assumed in analyzed accidents.

In its evaluation, NNECO concluded, based on its evaluation as

required by 10 CFR 50.92, that the proposed revision does not involve a

SHC.

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,

Connecticut.

NRC Section Chief: James W. Clifford.

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

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: May 13, 1999.

Description of amendment requests: The proposed amendments would

modify Technical Specification (TS) 6.2.A.2, ``Onsite and Offsite

Organizations,'' to reflect a change in the organizational structure

implemented on March 1, 1999.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

increase in the probability or consequences of an accident

previously evaluated.

The proposed change is administrative in nature and does not

significantly affect any system that is a contributor to initiating

events for previously evaluated accidents. Neither does the change

significantly affect any system that is used to mitigate any

previously evaluated accidents. Therefore, the proposed change does

not involve any significant increase in the probability or

consequence of an accident previously evaluated.

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

a new or different kind of accident from any accident previously

analyzed.

The proposed change is administrative in nature and does not

alter the design, function, or operation of any plant component nor

does the proposed change install any new or different equipment,

therefore the possibility of a new or different kind of accident

from those previously analyzed has not been created.

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

reduction in the margin of safety.

The proposed change is administrative in nature and does not

involve a significant reduction in the margin of safety associated

with the safety limits inherent in either the fuel cladding, RCS

[reactor coolant system] boundary, reactor containment, or other

structures, systems, or components (SSCs).

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: Claudia M. Craig.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: June 7, 1999.

Description of amendment request: The proposed amendments, if

approved, would revise Technical Specifications (TS) Section 3/4.4.3

and its associated TS Bases to reflect changes to refine and clarify

the action statement concerning inoperable reactor coolant leakage

detection systems.

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 proposed TS changes directly establish the minimum

acceptable level of Reactor Coolant System (RCS) leakage detection

instrumentation required to support plant power operations. The

level of RCS leakage detection capability inherent with the proposed

TS change will continue to provide acceptable early warning

detection of potential RCS pressure boundary degradation as required

under 10 CFR 50.36 (c)(2)(ii) (A) Criterion 1.

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

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes only affect systems associated with the

detection of accidents involving degradation of the RCS

[[Page 38035]]

pressure boundary. The proposed TS changes do not involve any

physical changes to plant structures, systems, or components. The

RCS Leakage Detection Systems will continue to function as designed

in all modes of operation. No new accident type is created as a

result of the proposed changes. No new failure mode for any

equipment is created. The changes are consistent with the guidance

provided in [Standard Technical Specifications General Electric

Plants BWR/4 dated April 1995] NUREG-1433, Revision 1, pertaining to

RCS Leakage Detection.

Therefore, the proposed TS 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 the margin of safety.

The TS Limiting Conditions for Operation (LCO) specify for

systems and equipment important to safety, the minimum level of

operability required to permit continued power operation. The

proposed TS changes revise this minimum level of operability by

permitting long term plant operation with the removal of the Drywell

Unit Coolers Condensate Flow Rate Monitoring System from service.

Currently, this condition would permit the plant to continue to

operate for up to 30 days. This change is not a reduction in the

margin of safety since:

The proposed Technical Specification LCO change for RCS Leakage

Detection Systems maintains four (4) diverse methods of detecting

RCS leakage and permits continuous operation with the Drywell Unit

Coolers Condensate Flow Rate Monitors out of service provided that

more frequent surveillance checks are provided for the containment

atmosphere monitoring system. The proposed TS change institutes the

additional surveillance requirements.

The LGS reactor coolant pressure boundary was designed to ASME

Class 1, Seismic Category I design criteria with no special

dispensation which would warrant such additional RCS leakage

detection capability or more stringent LCO criteria than those

generically approved under the Improved Standard Technical

Specifications.

Review of the TS Bases Section and UFSAR identified no

discussions regarding margin of safety for the RCS Leakage Detection

Systems, which would be reduced by the proposed Technical

Specification LCO change. It is further demonstrated that an

acceptable margin of safety exists based on the generic regulatory

approval of the Improved Standard Technical Specifications which

will remain bounded by the proposed LGS TS changes.

Therefore, the proposed TS changes do not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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

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

PA 19101.

NRC Section Chief: James W. Clifford.

Portland General Electric Company, Docket No. 50-344, Trojan Nuclear

Plant, Columbia County, Oregon

Date of amendment request: August 27, 1998.

Description of amendment request: The proposed amendment would

revise the Facility Operating (Possession-Only) License and the

Permanently Defueled Technical Specifications. Multiple license

conditions and technical specification requirements are proposed to be

deleted to reflect the transfer of the nuclear spent fuel from the

existing 10 CFR Part 50 licensed area to the 10 CFR Part 72 Independent

Spent Fuel Storage Installation (ISFSI) area.

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.

This proposed amendment reflects removal of the spent nuclear

fuel from the 10 CFR 50 licensed area and transfer of the spent

nuclear fuel to the 10 CFR 72 ISFSI licensed area. The probability

and consequences of accidents associated with storage of spent

nuclear fuel within the TNP [Trojan Nuclear Plant] ISFSI were

evaluated as part of PGE's 10 CFR 72 license application. Following

completion of the transfer of the spent nuclear fuel to the 10 CFR

72 licensed ISFSI and in light of the revised Appendix A Technical

Specification, Section 4.2, that precludes storage of spent nuclear

fuel within the 10 CFR 50 licensed area, the potential for accidents

associated with the storage and handling of fuel in the 10 CFR 50

licensed area will be eliminated. Therefore, deleting those

technical specifications associated with spent nuclear fuel will not

result in a significant increase in the probability or consequences

of accidents previously analyzed.

The proposed license amendment also relocates administrative

requirements from Section 5.0 of the Technical Specifications to

topical report PGE-8010, ``TNP Nuclear Quality Assurance Program.''

Relocation of administrative requirements follows the guidance

provided in NRC Administrative Letter 95-06. Relocating these

administrative requirements will not result in changes in method of

operation of any plant equipment, therefore these changes will not

result in a significant increase in the probability or consequences

of accidents previously evaluated.

The proposed license amendment will delete the on duty shift

manning requirements (Technical Specification 5.2.2a). With removal

of the spent nuclear fuel from the 10 CFR 50 licensed area, there

are no remaining important to safety systems required to be

monitored. With removal of the spent nuclear fuel from the 10 CFR 50

licensed area, there are no remaining credible accidents which

require the actions of a Shift Manager or non-certified operator to

prevent occurrence or mitigate consequences. Therefore, deleting the

shift manning requirements will not result in an increase in the

probability or consequences of an accident previously analyzed.

Deleting the Independent Review and Audit Committee (IRAC) is

also proposed in this license amendment request. The responsibility

of IRAC is to review and advise the Plant General Manager on matters

relating to the safe storage of irradiated fuel. Since approval of

this license amendment request is contingent upon removal of the

spent nuclear fuel from the 10 CFR 50 licensed area and a revised

Technical Specification Section 4.2 prevents future storage,

deleting IRAC will not result in an increase in the probability or

consequences of an accident previously evaluated.

This license amendment request proposes to revise and relocate

License Condition 2.C.(8), Fire Protection, to the TNP Quality

Assurance Program (PGE-8010). The revised text removes requirements

associated with making changes that could adversely impact the safe

storage of irradiated fuel. Following removal of the spent nuclear

fuel from the 10 CFR 50 licensed area and implementation of the

proposed revision to Technical Specification Section 4.2, irradiated

fuel will not be stored within the 10 CFR 50 licensed area so this

change will not result in an increase in the probability of

occurrence or consequences of accidents previously analyzed.

Relocation of the remaining requirements contained in this license

condition to the TNP Quality Assurance Program (PGE-8010) will

provide the necessary administrative control to ensure that changes

to the fire protection program will not increase the likelihood of

an offsite release of radioactive material due to a fire. Therefore,

this change will not result in an increase in the probability of

occurrence or consequence of accidents previously analyzed.

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

or different kind of accident from any accident previously

evaluated.

The proposed license amendment reflects the reduced operational

risks within the 10 CFR 50 licensed area after the spent nuclear

fuel has been transferred to the ISFSI. In addition, administrative

controls contained in Section 5.0 of the Technical Specifications

will [be] relocated to PGE-8010, TNP Nuclear Quality Assurance

Program. These changes have no impact on plant equipment and only an

administrative impact on some of the procedures used for operating

plant equipment, which may still be needed within the 10 CFR 50

licensed area following the

[[Page 38036]]

transfer of the spent nuclear fuel to the 10 CFR 72 ISFSI license

area. This proposed amendment does not result in the addition of new

equipment or result in the alteration of the operation of existing

structures, systems, or components. Therefore, the proposed changes

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

from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed 10 CFR 50 license amendment eliminates those

technical specifications and license conditions associated with the

storage of spent nuclear fuel. Following transfer of the spent

nuclear fuel to the 10 CFR 72 ISFSI, the potential for fuel related

accidents will be eliminated from the 10 CFR 50 licensed area.

Therefore, removal of those technical specifications and license

conditions associated with the safe storage of spent nuclear fuel

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

Relocating administrative programs in Technical Specification,

Section 5.0, ``Administrative Controls,'' follows the guidance of

NRC Administrative Letter 95-06. With the exception of deleting

those administrative controls associated with storage of spent

nuclear fuel, the administrative programs will be relocated to the

TNP Quality Assurance Program (PGE-8010). This administrative

relocation of requirements does not involve a significant reduction

in a margin of safety.

This proposed amendment also requests deleting several license

conditions and relocating License Condition 2.C.(8), ``Fire

Protection.'' The deleted license conditions were related to either

power operations or activities which have been completed and are no

longer required. Relocating License Condition 2.C.(8), ``Fire

Protection,'' to the TNP Quality Assurance Program (PGE-8010) will

continue to maintain the required level of administrative control

for the fire protection program since changes to PGE-8010 are

controlled in accordance with the requirements of 10 CFR

50.54(a)(3). Deleting these license conditions will, therefore, not

result in 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 10 CFR 50.92(c) are

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Branford Price Millar Library,

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

Portland, Oregon 97207.

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

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

NRC Section Chief: Michael T. Masnik.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: June 7, 1999, as supplemented by letter

dated June 24, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 2.0, Safety Limits and Limiting

Safety System Settings, TS 3.2.5, DNB [Departure from Nucleate Boiling]

Parameters, and the associated Bases, and Administrative Controls

Section 6.9.1.6, Core Operating Limits Report (COLR), by relocating

cycle-specific reactor coolant system-related parameter limits from the

TSs to the COLR. This would allow for flexibility to enhance plant

operating margin and/or core design margins without the need for cycle-

specific license amendment requests.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed amendment is a programmatic and administrative

change that does not physically alter safety-related systems, nor

does it affect the way in which safety-related systems perform their

functions. Because the design of the facility and system operating

parameters are not being changed, the proposed amendment does not

involve an increase in the probability or consequences of any

accident previously evaluated.

The cycle-specific limits in the Core Operating Limits Report

will continue to be controlled by the STP [South Texas Project]

programs and procedures. Each accident analysis addressed in the

UFSAR [Updated Final Safety Analysis Report] will be examined with

respect to changes in the cycle-dependent parameters, which are

obtained from the use of NRC-approved reload design methodologies,

to ensure that the transient evaluation of new reloads are bounded

by previously accepted analyses. This examination, which will be

conducted per the requirements of 10CFR50.59, will ensure that

future reloads will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The safety limits imposed in Technical Specification 2.1.1.1 and

2.1.1.2 are consistent with the values stated in the STP Updated

Final Safety Analysis Report. The Reactor Coolant System Flow value

in the Technical Specifications will be changed from the Minimum

Measured Flow to the Thermal Design System Flow (approved by the

Nuclear Regulatory Commission in Amendments 97 and 84 on September

29, 1998) consistent with WCAP-14483-P-A [`Generic Methodology for

Expanding Core Operating Limits Reports']. This change does not

involve an increase in the probability or consequences of any

accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

Removal of cycle specific variables has no influence or impact

on, nor does it contribute in any way to the probability or

consequences of an accident. No safety-related equipment, safety

function, or plant operation will be altered as a result of this

proposed change. The cycle specific variables are calculated using

the NRC-approved methods, and submitted to the NRC to allow the

staff to continue to trend the values of these limits. The Technical

Specifications will continue to require operation within the core

operating limits, and appropriate actions will be required if these

limits are exceeded. The safety limits imposed in Technical

Specification 2.1.1.1 and 2.1.1.2 are consistent with the values

stated in the STP Updated Final Safety Analysis Report. The Reactor

Coolant System Flow value in the Technical Specifications will be

changed from the Minimum Measured Flow to the Thermal Design Flow

(approved by the Nuclear Regulatory Commission in Amendments 97 and

84 on September 29, 1998) consistent with WCAP-14483-P-A. This

proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change does not involve a significant reduction in

a margin of safety.

The margin of safety is not affected by the removal of cycle

specific core operating limits from the Technical Specifications.

The margin of safety presently provided by current Technical

Specifications remains unchanged. Appropriate measures exist to

control the values of these cycle specific limits. The proposed

amendment continues to require operation within the core limits as

obtained from NRC-approved reload design methodologies, and the

actions to be taken if a limit is exceeded remain unchanged.

The development of the limits for future reloads will continue

to conform to those methods described in NRC-approved documentation.

In addition, each future reload will involve a 10CFR50.59 safety

review to assure that operation of the unit within cycle-specific

limits will not involve a significant reduction in the margin of

safety.

The safety limits imposed in Technical Specification 2.1.1.1 and

2.1.1.2 are consistent with the values stated in the STP Updated

Final Safety Analysis Report. The Reactor Coolant System Flow value

in the Technical Specifications will be changed from the Minimum

Measured Flow to the Thermal Design System Flow (approved by the

Nuclear Regulatory Commission in Amendments 97 and 84 on September

29, 1998) consistent with WCAP-14483-P-A. This proposed change does

not involve a significant reduction in the margin of safety.

[[Page 38037]]

The proposed amendment is a programmatic and administrative

change that provides assurance that plant operations continue to be

conducted in a safe manner. As stated previously, the proposed

amendment does not physically alter safety-related systems, nor does

it affect the way in which safety-related systems perform their

functions. Because the design of the facility and system operating

parameters are not being changed, 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

77488.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

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

NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of amendment request: June 3, 1999.

Description of amendment request: The proposed amendment would

modify the Technical Specifications to reduce the Allowable Value (Av)

used for Reactor Vessel Water Level--Low, Level 3 for several

instrument functions.

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.

The Reactor Vessel Water Level--Low, Level 3 functions are in

response to water level transients and are not involved in the

initiation of accidents or transients. Therefore, reducing the Level

3 Av does not increase the probability of an accident previously

evaluated. Additionally, the results of the safety evaluation

associated with the lowering of the Level 3 Av concludes that the

previously evaluated transient and accident consequences are not

significantly affected by the change. Therefore, the proposed

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment to lower the BFN Units 2 and 3 Reactor

Vessel Water Level--Low, Level 3 Av does not involve a hardware

change and the purpose of the Level 3 function is not affected. The

Level 3 functions will continue to fulfill their design objective.

Therefore, reduction of the Av does not result in the possibility of

a new or different kind of accident.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The results of the safety evaluation associated with the

reducing the BFN Units 2 and 3 Reactor Vessel Water Level--Low,

Level 3 Av concluded that transient and accident consequences remain

within the required acceptance criteria. Therefore, the margin of

safety is not reduced for any event evaluated.

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: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Sheri R. Peterson.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

Date of amendment request: June 23, 1999.

Brief description of amendments: The proposed license amendments

would change the way in which the Emergency Diesel Generator (EDG)

automatic trips are tested in Surveillance Requirement (SR) 3.8.1.13. A

note would also be added to specify the CPSES, Unit 2, test schedule in

SR 3.8.1.13.

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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The emergency diesel generators are used to support mitigation

of the consequences of an accident and are not considered to be

initiator of any previously analyzed accident. Revising the

surveillance to verify the bypass of non-critical EDG trips on both

LOOP [loss of offsite power] and SI [safety injection] separately

enhances the ability of the EDG to perform its safety function by

ensuring continued operation during DBAs [design-basis accidents].

Therefore, this change will not result in an increase in the

probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed change to the surveillance requirement involves an

EDG start circuit modification. The circuit modification has been

previously installed on Unit 2 during 2RF04 [CPSES Unit 2, fourth

refueling outage] for reasons other than the issue associated with

the FWLB [feedwater line break]. As a part of the Unit 2

installation a 50.59 evaluation was performed and it was determined

that the modification did not represent an unreviewed safety

question. The modification similar to Unit 2 will be implemented on

Unit 1 and therefore, as concluded in the safety evaluation for the

original modification, no new failure mechanisms will be introduced

by the proposed change. The EDGs are designed to provide electrical

power to equipment important to safety in the event of a loss of

offsite power. The proposed change to the SR enhances the confidence

that the EDGs will start and fulfill their safety related function.

Therefore, this change will not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

The proposed change will not alter any accident analysis

assumptions, initial conditions, or results. Revising the

surveillance requirement to verify the EDG trip bypass for the LOOP

and SI separately will enhance the confidence that the EDG starts as

assumed in the safety analyses and does not create any new failure

scenarios and no margin is reduced.

Therefore, this 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: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gram.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: April 16, 1999, as superseded on June 9,

1999.

[[Page 38038]]

Description of amendment request: The licensee proposed clarifying

the inservice inspection requirements for Vermont Yankee Nuclear Power

Station regarding the granting of relief from ASME Code requirements by

the NRC. The licensee also proposed changes to reflect the previous NRC

approval of the use of ASME Code Case N-560 at Vermont Yankee Nuclear

Power Station.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated?

This change is only an administrative change that: (1) clarifies

the NRC's authority to grant relief to a specific requirement, and

(2) conforms the TS language regarding GL 88-01 to agree with the

NRC's acceptance of ASME Code Case N-560 for use at VY. This

conclusion is justified in that:

(a) The pursuit of relief from the ASME code and the imposition

of alternative requirements are governed by 10CFR50.55a and require

NRC approval. There are several sections in the regulations under

which such relief can be granted. The removal of reference to a

specific section of CFR that may be used to grant relief has no

effect on plant equipment or its operation.

(b) Adding words to clarify the relationship between GL 88-01

and Code Case N-560 eliminates a contradiction in sample selection

criteria and does not affect any equipment or its operation.

These changes can be considered administrative in nature and do

not change any of the accident analyses for the facility. Thus,

there are no changes to the probability or consequences of accidents

previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

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

previously evaluated.

The revision of the wording in the TS to generalize the granting

of relief to the ASME code does not result in any changes to the

plant equipment or its operation. Similarly, adding words to allow

use of the NRC-approved alternative to the sample selection guidance

provided in GL 88-01 does not impact plant equipment or its

operation. These changes are administrative in nature and do not

result in the creation of any new or different kinds of accidents.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

This change primarily revises the wording in the TS to clarify

the NRC's authority to grant relief to ASME Section XI requirements.

The change maintains the requirement for NRC approval to be obtained

for such relief. Secondly, this change conforms the TS language

regarding GL 98-01 to agree with a previous relevant NRC disposition

[Reference (e)]. [The staff notes that reference (e) is an NRC

letter dated November 9, 1999, which approved the use of Code Case

N-560 at Vermont Yankee Nuclear Power Station.] These administrative

changes do not result in a reduction in any margin of safety.

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

this review, it appears that the three standards of 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: June 24, 1999

Description of amendment request: The amendment clarifies the basis

for the reactor protection system bypass of the turbine stop valve

(TSV) closure and turbine control valve (TCV) fast closure scram

signals at low power. The amendment clarifies that the analytical basis

for this bypass corresponds to a fraction of reactor rated thermal

power and not other measures of power, for instance, turbine power.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change clarifies the basis for the reactor

protection system bypass of the turbine stop valve closure and

turbine control valve fast closure scram signals. Consideration of

the bypass function itself only applies to certain pressurization

transients and not accident analyses.

The change properly states the basis for the scram bypass and

relates it to reactor thermal power and precludes potential

misinterpretation of the basis for the bypass setpoint. Turbine

power lags reactor power over the range of concern. Therefore,

changing terminology related to ``power'' to mean ``reactor power''

instead of ``turbine power'' is conservative. Accordingly, this

change can not be less restrictive.

The low power (TSV closure and TCV fast closure) scram signal

bypass does not initiate or mitigate any accident considered in the

Updated Final Safety Analysis Report. This function is enabled at

higher power to mitigate the effects of the pressurization transient

which results from TSV closure or TCV fast closure. This change will

not alter assumptions relative to the initiation or mitigation of

any accident event.

This change will not involve a significant increase in the

probability or consequences of an accident previously evaluated

since there is no physical alteration of the plant configuration or

relaxation of setpoints or operating parameters.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

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

previously evaluated.

The reactor protection system bypass of the turbine stop valve

closure and turbine control valve fast closure scram signals is not

considered an initiator of any accident. This change to clarify the

basis for applicability of the bypass does not create any new or

different kind of accident since it does not involve any change in

the physical configuration of the plant, nor relaxation of setpoints

or operating parameters.

VY has determined that the proposed change does not create the

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

previously evaluated because the change merely adds a more

restrictive interpretation to current terminology.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

The change involves reducing the potential for misinterpreting

the basis for the reactor protection system bypass of the turbine

stop valve closure and turbine control valve fast closure scram

signals and consequent potential for nonconservative operation of

the plant. As a result, the potential for operation of the plant in

an unsafe condition is reduced, thereby maintaining the margin of

safety.

VY has determined that the proposed change does not involve a

significant reduction in a margin of safety since operation of the

plant consistent with analytical bases of operation is further

assured.

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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

[[Page 38039]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Notice of Issuance of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: March 30, 1999.

Brief description of amendments: The amendments revised license

conditions in each of the operating licenses to delete those license

conditions that no longer apply, make an editorial change in the Unit 1

license, and provide clarifying information regarding the license

condition in each license concerning equalizer valve restrictions.

Date of issuance: June 25, 1999.

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 188 & 185.

Facility Operating License Nos. DPR-29 and DPR-30: The amendments

revised the licenses.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24195).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

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

Date of application for amendment: March 30, 1999.

Brief description of amendment: The amendment adds Section 4.0.2 to

allow a 24-hour grace period for performing inadvertently missed

surveillance.

Date of issuance: June 25, 1999.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 202.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 19, 1999 (64 FR

27317).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

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

Date of application for amendment: March 30, 1999.

Brief description of amendment: The amendment adds Section 4.0.2 to

allow a 24-hour grace period for performing inadvertently missed

surveillance.

Date of issuance: June 25, 1999.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 202.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 19, 1999 (64 FR

27317).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: August 6, 1998, as supplemented

by letter dated May 18, 1999.

Brief description of amendment: The amendment approves a change to

Technical Specification (TS) 3.1.3.2, ``Position Indicator Channels--

Operating,'' which adopts requirements that are consistent with NUREG-

1432, ``Standard Technical Specifications for Combustion Engineering

Plants.'' In addition, the amendment approves the relocation of TS

Table 3.8-1, ``Containment Penetration Conductor Overcurrent Protective

Devices,'' to licensee control procedures in accordance with the

guidance provided in Generic Letter 91-08, ``Removal of Component Lists

From Technical Specifications.''

Date of issuance: June 29, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days of issuance.

Amendment No.: 208.

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56245).

The May 18, 1999, letter provided clarifying information that did

not change the scope of the original application and the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 29, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 17, 1996, as supplemented by

letters dated October 22, 1998, and January 12 and February 5, 1999.

[[Page 38040]]

Brief description of amendment: The amendment extends the

surveillance test interval for the reactor trip circuit breakers from

monthly to quarterly and revises the appropriate Bases page.

Date of issuance: June 29, 1999.

Effective date: As of the date of issuance and shall be implemented

60 days from the date of issuance.

Amendment No.: 153.

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48261).

The October 22, 1998, and January 12 and February 5, 1999, letters

provided additional information that did not extend the scope of the

original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 29, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: October 27, 1998, supplemented

March 19, 1999.

Brief description of amendment: This amendment relocates a TS

surveillance requirement from TS Section /4.6.5.1, ``Shield Building--

Emergency Ventilation System'' to TS Section 3/4.6.5.2, ``Shield

Building Integrity.'' Administrative and bases changes have also been

made.

Date of issuance: June 22, 1999.

Effective date: June 22, 1999.

Amendment No.: 233.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64125). The March 19, 1999, supplement to the application did not

expand the scope of the original application as noticed, and did not

change the staff's proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: August 31, 1998, as revised on

March 18, 1999.

Brief description of amendment: The amendment approves changes to

the Improved Technical Specifications to allow a repair roll process

which would be used to repair steam generator tubes with defects within

the upper tubesheet. Changes to inservice inspection and reporting

requirements and several format and editorial changes were also

included.

Date of issuance: June 28, 1999.

Effective date: As of date of issuance, to be implemented prior to

commencing Cycle 12 operation.

Amendment No.: 179.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56249). The revised submittal dated March 18, 1999, expanded the scope

of the amendment request as originally noticed, and the application was

renoticed on April 21, 1999 (64 FR 19557).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 28, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: February 7, 1997, as

supplemented October 24, 1998

Brief description of amendment: The amendment incorporates changes

to more accurately reflect current plant design, adopts changes in

surveillance requirements consistent with the Standard Technical

Specifications, identifies changes to plant systems and revisions to

Technical Specifications system descriptions not involving Limiting

Conditions for Operations, and makes editorial or typographical

corrections.

Date of issuance: June 21, 1999.

Effective date: As of the date of issuance, and shall be

implemented within 30 days.

Amendment No.: 212.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 25, 1998 (63 FR

14486) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 21, 1999.

No significant hazards consideration comments received: No.

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.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: June 11, 1998.

Brief description of amendment: The amendment revises Technical

Specification 6.12.1 to allow use of an alternative high radiation area

control consistent with Regulatory Guide 8.38.

Date of issuance: July 1, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 213.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43204) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 1, 1999.

No significant hazards consideration comments received: No.

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.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336,

and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New

London County, Connecticut

Date of application for amendment: December 22, 1998, as

supplemented March 19, 1999.

Brief description of amendment: The amendment replaces specific

titles in Section 6.0 of the Technical Specifications of all three

Millstone units with generic titles.

Date of issuance: June 3, 1999.

Effective date: As of the date of issuance to be implemented within

30 days from the date of issuance.

Amendment No.: 105, 235, and 171.

[[Page 38041]]

Facility Operating License Nos. DPR-21, DPR-65, and NPF-49:

Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4158). The March 19, 1999, letter provided clarifying information that

did not change the scope of the December 22, 1998, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 3, 1999.

No significant hazards consideration comments received: No.

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.

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

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

Date of application for amendment: January 4, 1999, as supplemented

April 7, 1999.

Brief description of amendment: The amendment changes Technical

Specifications 3.5.2, ``Emergency Core Cooling Systems--ECCS

Subsystems--Tavg greater than or less than 300 deg.F;'' 3.6.2.1,

``Containment Systems--Depressurization and Cooling Systems--

Containment Spray and Cooling Systems;'' 3.7.1.2, ``Plant Systems--

Auxiliary Feedwater Pumps;'' 3.7.3.1, ``Plant Systems--Reactor Building

Closed Cooling Water System;'' and 3.7.4.1, ``Plant Systems--Service

Water System.'' The changes were made to the system pump flow

requirements to incorporate the results of revised hydraulic and

accident analyses.

Date of issuance: June 29, 1999.

Effective date: As of the date of issuance and shall implemented

within 60 days from the date of issuance.

Amendment No.: 236.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 14, 1999 (64 FR

2523).

The April 7, 1999, supplemental letter did not change the staff's

original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 29, 1999.

No significant hazards consideration comments received: No.

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.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: March 9, 1999, as supplemented May 3,

1999

Brief description of amendments: The amendments delete the

requirement to have an independent safety engineering group (ISEG) from

the Technical Specifications and applies the substantive requirements

now applicable to the ISEG to other organizations and relocates those

requirements from the Technical Specifications to Chapter 16 of the

Operational Quality Assurance Plan (OQAP). In the letter of May 3,

1999, the licensee submitted the changes to Chapter 16 of the OQAP to

incorporate the substantive Technical Specification requirements

currently applicable to the ISEG into the OQAP in the form of an

independent technical review program, and stated that these changes to

the OQAP will become effective upon approval of the amendments.

Date of issuance: June 23, 1999. Effective date: June 23, 1999, to

be implemented within 30 days. Implementation includes incorporating

the OQAP pages into the OQAP.

Amendment Nos.: Unit 1-112 ; Unit 2-99.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 7, 1999 (64 FR

17030) The May 3, 1999, supplement provided additional clarifying

information within the scope of the original notice and did not change

the staff's initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: May 27, 1999, as supplemented May 28,

1999.

Brief description of amendments: The amendments add a footnote to

Technical Specfiication 4.8.2.1e, ``D.C. Sources--Operating,'' which

would, on a one-time basis for Unit 1 Battery BT1ED2, allow TU Electric

to substitute a performance discharge test ``...in lieu of the battery

service test required by Specification 4.8.2.1d, twice within a 60

month interval.''

Date of issuance: June 28, 1999.

Effective date: As of the date of issuance.

Amendment Nos.: 65 and 65

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration (NSHC): Yes (64 FR 31881 dated June 14, 1999). The notice

provided an opportunity to submit comments on the Commission's proposed

NSHC determination. No comments have been received. The notice also

provided an opportunity to request a hearing by July 14, 1999, but

indicated that if the Commission makes a final determination, any such

hearing would take place after issuance of the amendments.

The May 28, 1999, letter provided clarifying information that did

not change the scope of the original application and the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments, finding of

exigent circumstances, and final NSHC determination are contained in

Safety Evaluation dated June 28, 1999.

Attorney for Licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, NW., Washington, DC, 20036.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

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

281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: February 16, 1999.

Brief Description of amendments: These amendments revise TS Section

4.2 for Units 1 and 2. The changes relax the surveillance requirements

for reactor coolant pump (RCP) flywheels. The flywheels provide

extended reactor coolant flow coastdown capability if electric power

for the RCPs is lost. Previously, the flywheel inspections included an

ultrasonic examination (UT) of areas of high stress

[[Page 38042]]

concentration at the base and keyway every 3 years, and complete UT

every 10 years. The changes require only a 10-year UT based upon an

analysis presented in a Westinghouse topical report which has been

reviewed and accepted by the NRC staff.

Date of issuance: July 1, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: 221 and 221.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

change the Technical Specifications.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24204). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act

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