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

Federal RegisterJan 28, 1998

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from January 5, 1998, through January 15, 1998.

The last biweekly notice was published on January 14, 1998 (63 FR

2271).

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 February 27, 1998, 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

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petition; and the Secretary or the designated Atomic Safety and

Licensing Board will issue a notice of a hearing or an appropriate

order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, 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, Docket No. 50-261, H.B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: December 17, 1997.

Description of amendment request: The requested amendment revises

Technical Specification Section 5.6.5, ``Core Operating Limits Report

(COLR).'' The revisions add reference to an additional approved

methodology for correlating departure from nucleate boiling (DNB)

ratios. The added methodology is the Siemens Power Corporation Topical

Report, EMF-92-153(P)(A), ``HTP: Departure from Nucleate Boiling

Correlation for High Thermal Performance Fuel.''

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 change adds a methodology that has been previously

reviewed and approved by the NRC for determining the DNB safety limit.

The new methodology utilizes the High Thermal Performance (HTP)

correlation developed by the fuel manufacturer, Siemens Power

Corporation. The HTP correlation is empirically based and results in a

DNB safety limit that corresponds to a 95% probability at a 95%

confidence level that DNB will not occur. The DNB ratio safety limit is

a conservative design value which is used as a basis for setting core

safety limits. The DNB correlation is not assumed to be an initiator of

analyzed events or transients, and use of the new DNB correlation will

not alter assumptions relative to mitigation of accident or transient

events. The proposed change has been confirmed to ensure that no

previously evaluated accident or transient results in a DNB less than

the DNB correlation safety limit. The HTP DNB correlation assures with

high confidence that, for accidents and transients that do not result

in a DNBR less than the HTP DNBR safety limit, departure from nucleate

boiling and subsequent fuel overheat will not occur in HTP fuel.

Therefore, the proposed change does not involve any 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 proposed change does not involve any physical alteration of

plant systems, structures, or components or changes in parameters

governing normal plant operation. The proposed change

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will allow use of the new DNB correlation in like manner as the

existing DNB correlation in the analysis of accidents and transients to

assure that the acceptance criteria for current analyses are met.

Therefore, the proposed change does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of

safety?

The proposed change allows use of a DNB correlation that determines

a safety limit that is slightly lower than the currently used DNB

correlation. While the slightly lower DNB correlation safety limit

allows a small increase in margin in analyzing accidents and

transients, the change from the existing DNB correlation to the

proposed DNB correlation is not directly comparable to the margin of

safety. This is because the margin of safety for a particular accident

or transient is that margin that results from the difference between

the DNBR calculated for the particular accident or transient using the

DNB correlation and the DNBR safety limit determined by the DNB

correlation. Since both the safety limit and the accident or transient

calculated DNB use the same DNB correlation, the margin of safety is

consistently calculated and evaluated for acceptability. Since both the

current and proposed DNB correlation closely approximate test data, and

they still meet the 95/95 criterion, and the new DNB correlation does

not result in a DNBR from an accident or transient less than the DNBR

correlation safety limit, the proposed change does not result in a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Gordon E. Edison, Acting.

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

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

Date of amendment request: June 6, 1997, as supplemented September

25, 1997.

Description of amendment request: The proposed amendment would

delete the requirement to sample the spray additive tank per Technical

Specification (TS) Table 4.1-2, ``Frequency for Sampling Tests,'' and

delete the sodium hydroxide (NaOH) reference in TS Section 5.2.C.1. The

request to delete the requirement and the reference was inadvertently

omitted as part of the licensee's original submittal dated August 22,

1996, supplemented March 28, 1997, to eliminate the requirement for the

NaOH containment spray additive and spray additive tank.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

evaluated?

Response:

The request to remove the requirement for the spray additive tank

was approved as part of Amendment No. 191 to Operating License No. DPR

26. By letter dated April 23, 1997, the Commission reviewed and

approved the amendment request. However, Consolidated Edison failed to

include the deletion of the requirement to sample the spray additive

tank. The removal of the requirement for the spray additive tank has

been analyzed and approved; therefore, there is no further basis for

continued testing of the tank. Further, the deletion of the requirement

would not involve a significant increase in the probability or

consequences of an accident previously evaluated.

(2) Does the proposed license amendment create the possibility of a

new or different kind of accident from any previously evaluated?

Response:

The proposed changes allow the containment safeguards to mitigate

the consequences of a design basis LOCA [loss-of-coolant accident] in a

manner equivalent to that previously approved. Therefore, the proposed

changes do not create an accident or malfunction of safety equipment of

a different type.

(3) Does the proposed amendment involve a significant reduction in

margin of safety?

Response:

With the proposed changes, all of the safety criteria previously

evaluated are still valid and remain conservative. Therefore, the

proposed amendment does not involve a significant reduction in the

margin of safety.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Director.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of amendment request: December 17, 1997.

Description of amendment request: The proposed amendments would

revise Section 6.9.1.9 of the Technical Specifications (TS) to

reference updated or recently approved topical reports, which contain

methodologies used to calculate cycle-specific limits contained in the

Core Operating Limits Report. These topical reports have all been

previously approved by the staff under licensing actions separate from

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

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

presented below.

1. Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed changes do not involve any modification to

existing systems, components, operating limits, or operating procedure.

Therefore, these proposed changes will have no impact on the

consequences or probabilities of any previously evaluated accidents.

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

kind of accident from any accident previously evaluated?

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No. No actual plant equipment or operating procedure will be

affected by the proposed changes. Hence, no new equipment failure modes

or accidents from those previously evaluated will be created.

3. Will the change involve a significant reduction in a margin of

safety?

No. Margin of safety is associated with confidence in the design

and operation of the plant. The proposed changes to the TS do not

involve any change to plant design or operation. Thus, the margin of

safety previously analyzed and evaluated is maintained.

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

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

determine that the amendment request involves no significant hazards

consideration.

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

Black Street, Rock Hill, South Carolina 29730.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina 28242-0001.

NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: December 17, 1997.

Description of amendment request: The proposed amendments would

revise Section 6.9.1.9 of the Technical Specifications (TS) to

reference updated or recently approved topical reports, which contain

methodologies used to calculate cycle-specific limits contained in the

Core Operating Limits Report. These topical reports have all been

previously approved by the staff under licensing actions separate from

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

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

presented below.

1. Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed changes do not involve any modification to

existing systems, components, operating limits, or operating procedure.

Therefore, these proposed changes will have no impact on the

consequences or probabilities of any previously evaluated accidents.

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

kind of accident from any accident previously evaluated?

No. No actual plant equipment or operating procedure will be

affected by the proposed changes. Hence, no new equipment failure modes

or accidents from those previously evaluated will be created.

3. Will the change involve a significant reduction in a margin of

safety?

No. Margin of safety is associated with confidence in the design

and operation of the plant. The proposed changes to the TS do not

involve any change to plant design or operation. Thus, the margin of

safety previously analyzed and evaluated is maintained.

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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina.

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,

422 South Church Street, Charlotte, North Carolina.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: September 23, 1997.

Description of amendment request: The proposed amendment changes

the Reactor Protective System and Engineering Safety Actuation System

trip set point and allowable values for steam generator low pressure.

The proposed amendment also relocates the RPS and ESFAS response time

tables from the Technical Specifications to the Safety Analysis Report.

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 and Accident Previously Evaluated.

The proposed changes included in this amendment request do not

affect the accident initiators in any of the accidents previously

evaluated. The proposed trip setpoints and allowable values for Steam

Generator Pressure--Low are being reduced by this proposed amendment

request. This change is necessary to increase the operating margin

between the full power steam generator pressure and these setpoints.

The change should reduce the probability of an inadvertent Main Steam

Isolation Signal (MSIS) from occurring at power since it will increase

the operating space between the operating pressure and the setpoints.

Therefore, this amendment request will not increase the probability of

any accident previously evaluated.

The secondary system pipe break safety analyses were reanalyzed for

the Steam Generator Pressure--Low setpoint reduction effort. This

effort included the removal of unnecessary analysis conservatisms

resulting in a significant reduction in the associated setpoints. The

proposed changes do not involve any change to the configuration or

method of operation of any plant equipment used to mitigate the

consequences of an accident. The previously evaluated accidents which

were determined to be impacted by this setpoint change were evaluated

with no significant increase in the consequences.

This amendment request contains the relocation of the Reactor

Protective System (RPS) and Engineered Safety Features Actuation System

(ESFAS) response time information from the Technical Specifications

(TS) to the Safety Analysis Report. This proposed change adopts the TS

``line-item improvement'' as recommended in NRC Generic Letter 93-08,

``Relocation of Technical Specification Tables of Instrument Response

Time Limits,'' dated December 29, 1993. The NRC has concluded that 10

CFR 50.36 does not require the response time tables to be retained in

TSs and has issued Generic Letter 93-08 as a line item improvement to

allow their removal. Response time testing will still be required by

the ANO-2 TS after the relocation of the associated response time

information in this amendment request. Relocating the response time

information for the RPS and ESFAS from the TS to the SAR will not alter

these surveillance requirements. Therefore, the relocated response time

portion of this amendment request is considered administrative in

nature and will not affect the probability or consequences of any

accident previously evaluated.

Therefore, this change does not involve a significant increase in

the

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probability or consequences of any accident previously evaluated.

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

Accident from and Previously Evaluated.

The proposed changes do not involve any physical modifications

(i.e., new systems, new components, etc.) to the plant. The proposed

changes do not involve any change to the configuration or method of

operation of any plant equipment used to mitigate the consequences of

an accident. The results of the accident reanalyzes suggest no

different phenomena or plant behavior than previously considered. The

Steam Generator Pressure Low setpoint change does not create any new or

different system actuations or interactions than evaluated previously.

The relocated response time portion of this amendment request is

considered administrative in nature and is not considered an accident

initiator. Therefore, this change does not create the possibility of a

new or different kind of accident from any previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The accidents which were determined to be impacted by the Steam

Generator Pressure Low setpoint change were evaluated to ensure

acceptable results are maintained. The instrument error calculations

supporting the lower Steam Generator Pressure Low setpoint and

allowable values will ensure the present accident analysis assumptions

are still maintained. The methodology used to determine the instrument

loop errors and uncertainties is the same as that used in previous

amendment requests that have been reviewed and approved by the NRC.

Based on these evaluations, the proposed changes do not involve a

significant reduction in a margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the reauested chance does not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

NRC Project Director: John Hannon.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: September 23, 1997.

Description of amendment request: The proposed amendment reduces

the minimum primary system flow that is specified in the technical

specifications to reflect the effects of increased primary system

resistance caused by steam generator tube plugging.

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.

Entergy Operations is proposing a change to the Technical

Specifications for Arkansas Nuclear One--Unit 2 (ANO-2) to accommodate

a larger number of plugged steam generator tubes. The proposed

amendment request will revise the Technical Specifications to

conservatively account for the reduced reactor coolant system (RCS)

flow effects of plugging up to 30 percent of the tubes in either steam

generator. This change will reduce the minimum RCS total flow rate from

120.4 x 106 lbm/hr to 108.4 x 106 lbm/hr until

the steam generators are replaced. The steam generators are currently

scheduled for replacement during the fall of the year 2000. After the

steam generators are replaced, the minimum RCS flow will then return to

the current value of 120.4 x 106 lbm/hr.

The tube plugs that are installed in the steam generators are

passive components by nature. This amendment request does not change

the type of plugs which may be installed in the steam generators nor

does it change the criteria for plugging steam generator tubes.

Reducing the minimum required RCS flow does not change the plant's

required mode of operation or modify any active component. Therefore,

this amendment request will not significantly increase the probability

of the occurrence of a previously evaluated accident.

The installation of steam generator tube plugs removes the affected

tube from service thus reducing the heat transfer surface area and

increasing the steam generator primary side flow resistance. The

increased flow resistance in the affected steam generator leads to a

reduction in the RCS flow available for core cooling. The reduced RCS

flow rate and heat transfer surface area resulted in a change in

several primary and secondary parameters that required reanalysis. The

ANO-2 accident reanalyses supporting the additional steam generator

tube plugging and the reduction in RCS flow have been completed.

The Design Basis Accidents (DBAs) affected by these changes were

reanalyzed to determine if the effects of increased steam generator

tube plugging and the reduced RCS flow could result in exceeding the

acceptance criteria applicable to each of these events. It was

determined that the DBA acceptance criteria would not be exceeded as a

result of increased steam generator tube plugging and reduction in the

minimum RCS flow rate.

Based on the results of the analysis, it is concluded that the

emergency core cooling system design satisfies the acceptance criteria

of 10 CFR 50.46(b) for a spectrum of small break and large break loss

of coolant accidents (LOCAs). The specified acceptable fuel design

limits (SAFDLs) and the RCS pressure boundary limits also are not

violated. The fuel and core performance were also determined to remain

within acceptable limits. Primary and secondary system pressures remain

below their respective pressure limits.

Analyses and evaluations of the DBAs have been performed

demonstrating that the NRC acceptance criteria for these events are

met. The revised analyses and evaluations consider reduced RCS flow,

increased RCS temperatures, and increased steam generator tube plugging

conditions. Although the offsite dose during a steam generator tube

rupture event could increase, the results remain well within 10 CFR

[Part] 100 limits. Therefore, the consequences of a previously

evaluated accident are not significantly increased.

Therefore, this change does not involve a significant increase in

the probability or consequences of any accident previously evaluated.

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

Accident from any Previously Evaluated.

The proposed amendment reduces the minimum RCS total flow to

account for the effects of steam generator tube plugging. This

amendment request will not change the modes of operation defined in the

Technical Specifications. This change does not add any new equipment,

modify any interfaces with any existing equipment, change the

equipment's function, or the method of operating the equipment. The

proposed change does not change plant conditions in a manner which

could

[[Page 4313]]

affect other plant components. Reactor core, RCS, and steam generator

parameters remain within appropriate design limits during normal

operation. The proposed change could not cause any existing equipment

to become an accident initiator. Therefore, this change does not create

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

previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The margins of safety associated with this change are defined in

the fuel and core related analyses, and in each of the transient and

accident analyses affected by the reduced RCS flow. An evaluation of

the affected analyses confirmed that the established acceptance

criteria for specified acceptable fuel design limits, primary and

secondary system over-pressurization, and the acceptance criteria for

the emergency core cooling systems have been satisfied by this license

amendment request. The evaluation concludes that, when considering the

proposed Limiting Conditions for Operation for the minimum RCS total

flow rate, all applicable acceptance criteria limits are met. The

margins of safety associated with the transient and accident analyses

affected by this change will not be significantly reduced. Therefore,

this change does not involve a significant reduction in the margin of

safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

NRC Project Director: John Hannon.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: December 10, 1997.

Description of amendment request: To clarify certain sections of

the Technical Specifications (TSs) and Bases which have been

demonstrated to be unclear or conflicting. Administrative changes

include TS 2.3 Bases, Table 3.1.1.G.1, Table 3.1.1.M.2, Section 4.3.C,

and Section 6.1.1. Technical changes include Table 3.3.3, note b,

Section 3.4 Bases, Section 3.8 Bases and Section 4.5 Bases.

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:

With respect to the administrative changes, they are typical of the

example I.c.2.e.i in 51 FR 7744 and therefore, they do 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 the margin of safety; in that

they are purely administrative changes to achieve consistency or

correct an error in the TS.

With respect to technical change, Table 3.1.1, note b:

1. Involve a siginificant increase in the probability or

consequences of an accident previously evaluated; (or)

The proposed change would restore the original value of less than

600 psig. This lower value would not increase the probability of any

accident as it provides a more conservative level below which

protection can be bypassed.

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

from any accident previously evaluated; (or)

The proposed change would restore the original value of less than

600 psig. The setpoint of a bypass cannot create a different kind of

accident, it can only affect the severity.

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

requested change lowers the bypass setpoint, the margin of safety will

be increased.

With respect to Section 3.4 Bases:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated; (or)

The proposed change to the Bases removes a possible area of

confusion from the [TS], and updates the Bases to reflect the results

of newer, approved methodologies. Therefore, no change to any

probability calculation occurs.

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

from any accident previously evaluated; (or)

The proposed change addresses an existing accident (Small Break

LOCA) and removes outdated and possibly confusing information.

Therefore, no new or different kind of accident is created.

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

The proposed change does not change the way the plant is operated

or the way design Bases are maintained. It only removes an outdated and

possibly confusing paragraph from the Bases, therefore, no margin of

safety is affected.

With respect to Section 3.8 Bases:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated; (or)

The Isolation Condenser Radiation Monitors had no impact o[n] the

operation of any plant system. Additionally, the monitors were not

relied upon for any post accident evaluations. They were removed from

the plant using the 10 CFR 50.59 process. As this request updates the

[TS] Bases to reflect the plant as currently configured, no impact on

the probability or consequences of any previously evaluated accident is

possible.

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

from any accident previously evaluated; (or)

The Isolation Condenser Radiation Monitors had no impact o[n] the

operation of any plant system. Additionally, the monitors were not

relied upon for any post accident evaluations. They were removed from

the plant using the 10 CFR 50.59 process. As this request updates the

[TS] Bases to reflect the plant as currently configured, no new or

different kind of accident is created.

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

The Isolation Condenser Radiation Monitors had no impact o[n] the

operation of any plant system. Additionally, the monitors were not

relied upon for any post accident evaluations. They were removed from

the plant using the 10 CFR 50.59 process. As this request updates the

[TS] Bases to reflect the plant as currently configured, no reduction

in any margin of safety can occur.

With respect to Section 4.5 Bases:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated; (or)

No change to any procedure, nor any modification to any system is

requested. The same surveillance will be performed at the same

frequency. Only the brand of chemical used to perform the surveillance

will be affected. As an equivalent chemical will be selected, no

increase in the probability or consequences of an accident previously

evaluated can be created.

[[Page 4314]]

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

from any accident previously evaluated; (or)

No change to any procedure, nor any modification to any system is

requested. The same surveillance will be performed at the same

frequency. Only the brand of chemical used to perform the surveillance

will be affected. As an equivalent chemical will be selected, no new or

different kind of accident previously evaluated can be created.

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

No change to any procedure, nor any modification to any system is

requested. The same surveillance will be performed at the same

frequency. Only the brand of chemical used to perform the surveillance

will be affected. As an equivalent chemical will be selected, no margin

of safety can be affected.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

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

NRC Project Director: Ronald B. Eaton.

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

Linn County, Iowa

Date of amendment requests: October 3, 1997.

Description of amendment requests: The proposed amendment would

revise the Operating License to allow the start of core offload as soon

as 60 hours after shutdown instead of the 120 hours currently

specified.

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 Operating License Amendment will not significantly

increase the probability or consequences of any previously evaluated

accidents.

The proposed change will allow initiation of core offload earlier

after shutdown than is currently allowed. Thermal-hydraulic analysis

shows that maximum bulk SFP, local water, and fuel clad temperatures

will remain within acceptable limits and, in fact, do not exceed those

previously reviewed and approved for Amendment 195.

Thermal-hydraulic analysis shows the minimum time to action is

calculated at 4.5 hours versus 5.5 hours previously reviewed and

approved for Amendment 195. In the event of a loss of forced cooling

with cask pit isolation gate failure event, the DAEC will use Emergency

Service Water (ESW), a Seismic Category I system, to provide makeup to

the SFP. It is estimated to take no more than 2 hours to provide ESW

makeup to the SFP, therefore the minimum time to action of 4.5 hours is

sufficient time to prevent uncovering the fuel in the SFP.

The DAEC design basis refueling accident, as discussed in Section

15.10.2 of the Updated Final Safety Analysis Report, assumes a twenty-

four hour decay time before core offload begins. The proposed change

does not adversely affect that accident analysis.

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

probability or consequences of an accident previously evaluated.

2. The proposed changes will not create a new or different kind of

accident from those previously evaluated.

Thermal-hydraulic analysis shows that the proposed change will not

result in maximum bulk SFP, local water, or fuel clad temperatures

which would initiate bulk pool boiling, challenge fuel rod integrity or

jeopardize the structural integrity of the pool.

As stated above, the minimum time to action of 4.5 hours allows

sufficient time to provide ESW makeup to the SFP. Therefore, this

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

accident.

3. The proposed change will not result in a significant reduction

in any margin of safety.

This change will not result in maximum bulk SFP, local water, and

fuel clad temperatures in excess of those previously evaluated and

accepted per Amendment 195. The thermal-hydraulic analysis for Case C

does show a reduction in the minimum time to action by one hour.

However, 4.5 hours does provide sufficient time to provide ESW makeup

to the SFP as this task is estimated to require no more than 2 hours.

Furthermore, this change does not result in any change to the Technical

Specifications. Therefore, this change does not result in a significant

reduction in a margin of safety.

Based upon the above, we have determined that the proposed

amendment will not involve a significant hazards consideration.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: December 15, 1997.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) 2.1 and 3/4.4.1 to change the

safety limit minimum critical power ratio (MCPR) for the upcoming fuel

operating cycle (Cycle 7) from 1.07 to 1.09 for two recirculation loop

operation and from 1.08 to 1.10 for single loop operation. An obsolete

footnote in TS 3/4.4.1, which states that ``the MCPR Safety Limit of

1.07 will be used through the first operating cycle,'' would be

deleted. The associated Bases 2.1 would be changed to (1) reflect the

new MCPR values, (2) delete certain details (including Bases Table

B2.1.2-1, ``Uncertainties Used in the Determination of the Fuel

Cladding Safety Limit,'' and Bases Table B2.1.2-2, ``Nominal Values of

Parameters Used in the Statistical Analysis of Fuel Cladding Integrity

Safety Limit,'') and (3) substitute for the deleted detail a reference

to General Electric Standard Application for Reactor Fuel (GESTAR II),

NEDE-24011-P-A, and to the cycle-specific analysis. The TS Index would

be changed to reflect deletion of Bases Tables B2.1.2-1 and B2.1.2-2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The derivation of the revised Safety Limit MCPR was performed using

the NRC approved methodology in GESTAR II. The Safety Limit MCPR is a

TS numerical value that cannot initiate an event. Maintaining

compliance with this

[[Page 4315]]

limit will assure that 99.9 percent of the fuel rods will not

experience transition boiling during transient events. The deletion of

the footnote that is no longer necessary and the revision to the Bases

information are administrative only. The proposed change does not

modify any of the accident initiators described in the USAR [Updated

Safety Analysis Report]. No equipment malfunctions or procedural errors

are created as a result of this change, therefore, no accidents are

affected by it. The change does not adversely impact the integrity of

the fuel cladding, which is the first barrier to the release of

radioactivity to the environment. The change does not affect the

operation of any systems necessary to mitigate the radiological

consequences of an accident or to safely shutdown the plant. Therefore,

this change will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

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

different kind of accident from any accident previously evaluated.

The Safety Limit MCPR is a TS numerical value designed to prevent

fuel damage from transition boiling. It cannot create the possibility

of a transient or accident. The deletion of the footnote that is no

longer necessary and the revision to the Bases information are

administrative only. The proposed change does not directly impact the

operation of any systems or equipment important to safety. The analyses

show that all fuel licensing acceptance criteria are met. The fuel

cladding, reactor vessel, and reactor coolant system integrity will be

maintained. Therefore, this change will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

The Safety Limit MCPR calculation was performed using the NRC

approved methodology in GESTAR II. Analyses of limiting USAR transients

establish Operating Limit MCPR values that ensure that the Safety Limit

MCPR is not violated. The revised cycle specific Safety Limit MCPR

preserves the existing margin of safety and will continue to assure

that 99.9 percent of the fuel rods will not experience transition

boiling during transient events. The deletion of the footnote that is

no longer necessary and the revision to the Bases information are

administrative only. Thus, the margin of safety to fuel cladding

failure due to insufficient cladding heat transfer during transient

events is not reduced. Therefore, this change will not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: S. Singh Bajwa.

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

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

Date of amendment request: November 13, 1997.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) to (1) modify the low

temperature overpressure protection (LTOP) requirements; (2) modify the

reactor coolant system (RCS) heatup and cooldown limits; and (3) make

changes to correct various items based on the licensee's review of the

current TSs. The supporting TS Bases sections would also be changed to

reflect the proposed TS changes.

The affected TSs are: TS 3.1.2.1, ``Flow Paths--Shutdown;'' TS

3.1.2.2, ``Flow Paths--Operating;'' TS 3.1.2.3, ``Charging Pump--

Shutdown;'' TS 3.1.2.4, ``Charging Pumps--Operating;'' TS 3.1.2.5,

``Boric Acid Pumps--Shutdown;'' TS 3.1.2.6, ``Boric Acid Pumps--

Operating;'' TS 3.1.2.8, ``Borated Water Sources--Operating;'' TS

3.4.1.3, ``Coolant Loops and Coolant Circulation--Shutdown;'' TS 3.4.3,

``Relief Valves;'' TS 3.4.9.1, ``Reactor Coolant System;'' TS 3.4.9.2,

``Pressurizer;'' TS 3.4.9.3, ``Overpressure Protection Systems;'' TS

3.5.3, ``ECCS Subsystems--Tavg 2.2 in2 has been

established. This curve and rate may also be used when RCS cold leg

temperature is below 230 deg.F to demonstrate compliance with Appendix

G limits when unanticipated temperature excursions occur.

The current action statements of Technical Specification 3.4.9.1

will be separated by Mode and will be modified. Similar changes will be

made to the action statements of Technical Specification 3.4.9.2. A

time limit of 72 hours will be placed on the performance of the

engineering evaluation. If this evaluation is not performed in this

time period, or the evaluation does not allow continued operation, the

plant will be required to enter Mode 5 ([less than or equal to] 200

deg.F), instead of the current requirement to be q;''

(2) correct the wording in TS 4.1.1.1.2 ``Reactivity Control Systems

Shutdown Margin--Tavg [less than or equal to] 200 deg.F;''

(3) correct the mode applicability from Mode 3 to Modes 1 and 2 in TS

3.1.3.4 ``Reactivity Control Systems--Rod Drop Time;'' (4) correct the

terminology used to refer to the power dependent insertion limit alarm

in TS 4.1.3.6 ``Reactivity Control Systems--Regulating CEA [Control

Element Assembly] Insertion Limits;'' (5) add a footnote for Mode 4

operability requirement clarification to TS 3.5.3 ``Emergency Core

Cooling Systems, ECCS Subsystems--Tavg q) used to verify compliance

with Technical Specification 3.2.4 is associated with the quadrant of

highest power production with respect to the average of the four

quadrants, instead of the quadrant that deviates the most (increases or

decreases) from the average of the four quadrants. This is consistent

with the method by which power distribution factors are calculated and

applied in the accident analysis and how the Core Power Distribution

Monitoring System calculates Tq. The proposed change will

not alter the way Tq is calculated by the Core Power

Distribution Monitoring System, nor will it alter any of the power

distribution assumptions used in the accident analysis. Therefore, this

change will not significantly increase the probability or consequences

of an accident previously evaluated.

Surveillance Requirement (SR) 4.1.1.1.2 requires that the

difference between predicted and measured core reactivity values be

maintained within [plus or minus] 1.0% [delta]k/k, and that an

adjustment be made between the measured and predicted core reactivity

conditions prior to exceeding 60 EFPD [effective full power days]

following a refueling outage. The proposed change will not affect the

requirement to maintain predicted and measured core reactivity values

within [plus or minus] 1.0% [delta]k/k. However, it will no longer be

necessary to make an adjustment prior to exceeding 60 EFPD provided the

[plus or minus] 1.0% [delta]k/k requirement is met. Historically, this

difference has been small at Millstone Unit No. 2 (less than

approximately [plus or minus] 1.0% [delta]k/k) and an adjustment has

not been necessary to ensure the [plus or minus] 1.0% [delta]k/k

requirement is met. The fact that no adjustment (normalization) will be

necessary when reactivity differences are small will not affect the

ability to identify reactivity anomalies. Therefore, this change will

not significantly increase the probability or consequences of an

accident previously evaluated.

The proposed change to Technical Specification 3.1.3.4 will change

the applicability from Mode 3 to Modes 1 and 2. This is necessary to

allow performance of SR 4.1.3.4 at the conditions in the accident

analysis, and also specified in the [Limiting] Condition [for]

Operation (LCO). CEA [Control Element Assembly] drop time is important

for the mitigation of accidents that are initiated while the reactor is

critical. To ensure the CEA drop time assumed in the accident analysis

is valid, it is necessary to verify CEA drop time with plant conditions

consistent with those expected when the reactor is critical. This

proposed change will allow this verification, and thereby ensure the

CEAs will function as designed to mitigate design basis accidents.

Therefore, this change will not significantly increase the probability

or consequences of an accident previously evaluated.

The proposed change to SR 4.1.3.6 will modify the terminology used

to refer to the power dependent insertion limit (PDIL) alarm to agree

with plant terminology. This change will not alter equipment operation

or any technical aspect of the SR. The information added to the Bases

will specify what equipment provides the PDIL alarm. These changes will

eliminate any confusion with alarm terminology. Therefore, this change

will not significantly increase the probability or consequences of an

accident previously evaluated.

Technical Specification 3.5.3 requires an operable flowpath capable

of taking a suction from the refueling water storage tank (RWST) on a

safety injection actuation signal (SIAS), and automatically

transferring suction to the containment sump on a sump recirculation

actuation signal (SRAS) in Mode 4. In Mode 4, the automatic SIAS

generated by low pressurizer pressure and high containment pressure,

and the automatic SRAS generated by low RWST level, are not required to

be operable. Automatic actuation in Mode 4 is not required because

adequate time is available for plant operators to evaluate plant

conditions and respond by manually operating engineered safety

[[Page 4320]]

features components. Since the manual actuation (trip pushbuttons)

portions of the safety injection and sump recirculation actuation

signal generation are required to be operable in Mode 4, credit can be

taken for remote manual operation to generate the SIAS and SRAS which

will position all components to the required accident position. The

proposed change to Technical Specification 3.5.3 will add a footnote

(***) to explain how these requirements are met in Mode 4. This change

will not reduce operability or surveillance requirements for the

Emergency Core Cooling System (ECCS) subsystem required to be operable

by Technical Specification 3.5.3. The ECCS will continue to function as

designed to mitigate design basis accidents. Therefore, this change

will not significantly increase the probability or consequences of an

accident previously evaluated.

The proposed change to Technical Specification 3.6.3.2 will revise

the wording of the LCO and SR by changing ``locked closed'' to ``sealed

closed,'' and deleting the requirement to be electrically deactivated.

The action statement will also be revised to reflect these proposed

changes. These changes will not affect the requirement for the

containment purge valves to be closed in Modes 1 through 4. Therefore,

the proposed changes will not significantly increase the probability or

consequences of an accident previously evaluated.

The proposed change to SR 4.6.1.7 will change the surveillance

frequency from ``prior to each reactor startup'' to ``at least once per

31 days.'' This change, which will require the surveillance to be

performed more often (assuming a normal plant startup sequence) will

provide additional assurance that the containment purge valves are

sealed closed. In addition, this change will ensure consistency between

the SR and the applicability of this specification, and also with the

requirements to verify containment integrity in accordance with

Technical Specification 3.6.1.1. Therefore, the proposed change will

not significantly increase the probability or consequences of an

accident previously evaluated.

The change in numbering of SR 4.6.1.7 to SR 4.6.3.2 is an

administrative change only. It will not affect any technical aspect of

the SR. Therefore, the proposed change will not significantly increase

the probability or consequences of an accident previously evaluated.

The proposed changes to Technical Specifications 3.8.2.1 and

3.8.2.1A will modify the nomenclature used to refer to the vital A.C.

buses to be consistent with the terminology used by Operations

Department personnel and the nomenclature contained in their

procedures. These changes will not alter equipment operation or any

technical aspects of these specifications. These proposed changes are

administrative changes only. The A.C. buses will continue to function

as designed to mitigate design basis accidents. Therefore, these

changes will not significantly increase the probability or consequences

of an accident previously evaluated.

The proposed changes to Technical Specifications 5.1.3, 5.2.3,

5.3.2, and 5.7.1 will remove the word ``original.'' Reference to

original design is not appropriate since these items can be changed by

approved processes. However, these changes will still require the items

addressed by these specifications to be designed and maintained in

accordance with the Final Safety Analysis Report (FSAR). The proposed

changes have no affect on the current approved plant design. Therefore,

these changes will not significantly increase the probability or

consequences of an accident previously evaluated.

Technical Specification 5.9 will be deleted. The required

provisions for shoreline protection have been completed, and this

Technical Specification is no longer necessary. The removal of this

outdated specification will not impact any current requirements.

Therefore, this change will not significantly increase the probability

or consequences of an accident previously evaluated.

The proposed changes do not alter how any structure, system, or

component functions. There will be no effect on equipment important to

safety. The proposed changes have no effect on any of the design basis

accidents previously evaluated. Therefore, this License Amendment

Request does not impact the probability of an accident previously

evaluated, nor does it involve a significant increase in the

consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no new

or different type of equipment will be installed) or require any new or

unusual operator actions. They do not alter the way any structure,

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

plant is operated. The proposed changes do not introduce any new

failure modes. Therefore, the proposed changes will not create the

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

previously evaluated.

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

The proposed change to the definition of Tq will make

the Technical Specification definition consistent with the approved

calculation methodology. This will ensure the core power distribution

is consistent with accident analysis assumptions. The proposed change

to the wording of SR 4.1.1.1.2 will not affect the acceptance criteria

of [plus or minus] 1.0% [delta]k/k, which ensures the accident analysis

accurately reflects core reactivity conditions. The proposed change in

the applicability of Technical Specification 3.1.3.4 will allow

verification of CEA drop time at plant conditions assumed in the

accident analysis. This will ensure the CEAs will function as assumed.

The proposed change to SR 4.1.3.6 will modify the terminology used to

refer to the PDIL alarm to agree with plant terminology. This change

will not alter equipment operation or any technical aspect of the SR.

Adding the footnote to Technical Specification 3.5.3 will not change

any technical aspects of this specification. One ECCS subsystem will be

available for accident mitigation. The proposed change in wording of

Technical Specification 3.6.3.2 will not affect the requirement for the

containment purge valves to be closed in Modes 1 through 4. The

proposed change in the frequency of performance for SR 4.6.1.7 will

provide greater assurance that the containment purge valves are closed

to prevent the potential release of radioactive material through these

penetrations during accident conditions. The proposed changes in

terminology in Technical Specifications 3.8.2.1 and 3.8.2.1A will not

change any technical requirements for the equipment covered. The

equipment will still function as assumed. Modifying the Bases of

Technical Specifications are necessary to be consistent with the

proposed changes will not change any requirements of these

specifications. The modification to Technical Specifications 5.1.3,

5.2.3, 5.3.2, and 5.7.1 will not affect the requirement to maintain

these items in accordance with requirements contained in the FSAR.

Deleting Technical Specification 5.9 will not affect any requirements

since the requirements contained in this specification have already

been completed.

The proposed changes do not affect any of the assumptions used in

the accident analysis, nor do they affect any operability requirements

for equipment

[[Page 4321]]

important to plant safety. Therefore, these proposed changes will not

result in a significant reduction in the margin of safety as defined in

the Bases for Technical Specifications covered in this License

Amendment Request.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: December 12, 1997.

Description of amendment request: The proposed amendment would

revise the facility Technical Specifications (TSs) regarding normal

working hours of plant staff to provide for shift duration of 12 hours.

It would also revise the TSs to maintain existing ``once per shift''

surveillance requirements at 8-hour intervals.

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 proposed licensing amendment involve a significant

increase in the probability or consequences of an accident previously

evaluated?

Response:

Establishing operating personnel work hours at ``a normal 8 to 12

hour day, nominal 40-hour week'' allows normal plant operations to be

managed more effectively and does not adversely affect performance of

operating personnel. Overtime remains controlled by site administrative

procedures in accordance with NRC Policy Statement on working hours

(Generic Letter 82-12). If 8 hour shifts are maintained in part or

whole, then acceptable levels of performance from operating personnel

is assured through effective control of shift turnovers and plant

activities. No physical plant modifications are involved and none of

the precursors of previously evaluated accidents are affected.

Therefore, this change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c.

without changing the intent or meaning. [...] Changes to sections

4.5.F.3., 4.5.F.4., 4.5 Bases, and 4.7.A.7.a. do not change the intent

or meaning of the Technical Specifications, do not change operating

procedures, and are consistent with surveillance requirements.

[Therefore, the proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.]

Does the proposed license amendment create the possibility of a new

or different kind of accident from any accident previously evaluated?

Establishing operating personnel work hours at ``a normal 8 to 12

hour day, nominal 40-hour week'' allows normal plant operation to be

managed more effectively and does not adversely affect performance of

operating personnel. If 8 hour shifts are maintained in part or whole,

then acceptable levels of performance from operating personnel is

assured through effective control of shift turnovers and plant

activities. Overtime remains controlled by site administrative

procedures in accordance with the NRC Policy Statement on working hours

(Generic Letter 82-12). No physical modification of the plant is

involved. As such, the change does not introduce any new failure modes

or conditions that may create a new or different accident. Therefore,

plant operation in accordance with the proposed amendment will not

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

previously evaluated.

Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c.

without changing the intent or meaning. [* * *] Changes to sections

4.5.F.3., 4.5.F.4., 4.5 BASES, and 4.7.A.7.a. do not change the intent

or meaning of the Technical Specifications or operating procedures. All

previously performed functions are being maintained.

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

different kind of accident from any accident previously evaluated.

Does the proposed amendment involve a significant reduction in a