4308 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

the working papers at the Website,

select ‘‘Final Rule on Radiological

Criteria for License Termination,’’ then

select ‘‘Lic Term Document Library,’’

then select ‘‘Regulatory Guide,’’ and

then select ‘‘Module C.2: Regulatory

Position—Final Status Survey,’’ or

‘‘Module C.1: Regulatory Position—Dose

Modeling.’’

Meeting Agenda

9:00 Welcome and introduction

9:05 Presentation describing issues

considered in developing the draft

working paper

10:30 Break

10:45 Public comments on the draft

working paper. Attendees will be

asked for questions and comments

on each section of the draft working

paper.

12:00 Lunch

1:30 Continuation of public comments.

5:00 Adjourn

Submitting Written Comments

Comments may be posted

electronically on the NRC Technical

Conference Forum Website mentioned

above. Comments submitted

electronically can also be viewed at that

Website. Comments may also be mailed

to the Chief, Rules and Directives

Branch, U.S. Nuclear Regulatory

Commission, Washington, DC 20555.

FOR FURTHER INFORMATION: For

information or questions on meeting

arrangements, contact Nina Barnett,

Office of Nuclear Regulatory Research,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555, telephone 301–

415–6187, fax 301–415–5385, E-mail:

NMB@NRC.GOV. For technical

information or questions, contact

Stephen A. McGuire, Office of Nuclear

Regulatory Research, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, telephone 301–415–6204;

fax: 301–415–5385; E-mail:

SAM2@NRC.GOV.

Dated at Rockville, Maryland this 22nd day

of January, 1998.

For the Nuclear Regulatory Commission.

Cheryl Trottier,

Chief, Radiation Protection and Health Effects

Branch, Division of Regulatory Applications,

RES.

[FR Doc. 98–2017 Filed 1–27–98; 8:45 am]

BILLING CODE 7590–01–P

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
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4309

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
4310 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

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?
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4311

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
4312 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

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 120.4×106 lbm/

hr to 108.4×106 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

120.4×106 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
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 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 overpressurization, 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.
4314 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

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
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 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 < 300 °F;’’ and TS

3.10.3, ‘‘Pressure/Temperature

Limitation—Reactor Criticality.’’

The November 13, 1997, submittal

provides specific details related to each

of the proposed 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. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

Each of the proposed changes have

been grouped together, as appropriate,

to address this criteria.

HPSI Pump Not Required To Be

Operable In Modes 5 and 6. The

proposed change to only require one

charging pump to be operable in Modes

5 and 6, instead of the current

requirement for one charging pump and

one high pressure safety injection pump

(HPSI) pump to be operable, will result

in sufficient, but not excessive, Reactor

Coolant System (RCS) makeup

capability. When the plant is in Mode

5 or 6 there are two major factors to

consider with respect to the number of

RCS makeup pumps required to be

operable. If too many RCS makeup

pumps are required, an inadvertent start

of these pumps can result in a mass

addition transient beyond the capacity

of the Low Temperature Overpressure

Protection (LTOP) System. This may

result in an RCS pressure increase that

exceeds the 10CFR50 Appendix G

pressure/temperature limits.

Compliance with the mass input and

venting restrictions contained in the

proposed Technical Specification

3.4.9.3 will ensure the Appendix G

limits are not exceeded.

The minimum number of RCS

makeup pumps required to be operable

in Modes 5 and 6 ensures sufficient

makeup capability is available for RCS

inventory control and RCS boration

requirements. RCS inventory control is

necessary in Modes 5 and 6 to ensure

sufficient water is available for core

cooling. A rapid loss of RCS inventory

due to catastrophic pipe failures is

unlikely in Modes 5 and 6 due to the

reduced RCS pressure and temperature.

An inventory loss is more likely to

occur due to small system component

failures or during infrequently

performed evolutions, such as reduced

inventory operation. This type of

inventory loss will occur at a slower

rate. Plant operators will have time to

perform the necessary actions to

mitigate the event. Reliance on

automatic operation of the Emergency

Core Cooling System is not necessary

and Technical Specifications do not

require automatic actuation by the

Engineered Safety Features Actuation

System to be operable in Mode 4 or

below. Operator action is sufficient to

mitigate a loss of RCS inventory in

Mode 4 or below, provided sufficient
4316 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

RCS makeup capability is available.

Plant procedures and shutdown risk

management will provide adequate

administrative control to ensure

sufficient RCS makeup capability is

available, or that contingency plans

have been developed.

The minimum number of RCS

makeup pumps required to be operable

in Modes 5 and 6 ensures sufficient

makeup capability is available for RCS

boration requirements. The RCS is

required to be borated to a sufficient

boron concentration to ensure the

Technical Specification Shutdown

Margin (SDM) requirements are met.

The appropriate SDM requirements

must be met before entry is allowed into

Mode 5 or 6. RCS boron concentration

is increased to establish the required

SDM. This is normally accomplished by

adding borated water to the RCS during

plant cooldown to compensate for the

contraction of the RCS inventory. The

proposed change will restrict the

number of pumps available, which will

increase the time required to adequately

borate the RCS. However, the change

will not affect the ability to add boric

acid to the RCS.

Even though the proposed change will

remove the Technical Specification

requirement for an operable HPSI pump

in Modes 5 and 6, sufficient RCS

makeup capability will be available to

meet RCS boration and inventory

requirements. Therefore, the proposed

change will not result in a significant

increase in the probability or

consequences of an accident previously

evaluated.

LTOP Mass Input and Vent Size

Requirements. The proposed changes to

the RCS venting requirements currently

contained in Technical Specification

3.4.9.3, and the RCS makeup

requirements that will be relocated to

Technical Specification 3.4.9.3 are

necessary to be consistent with the new

LTOP analysis. These changes will

ensure the 10CFR50 Appendix G limits

are not exceeded.

The proposed changes to the mass

input restrictions will still allow two

charging pumps and one HPSI pump to

be capable of injecting into the RCS

when the RCS is operating in Mode 4

[less than or equal to] 275 °F. This

combination will be allowed in Mode 5

until RCS temperature is [less than or

equal to] 190 °F. When RCS cold leg

temperature is at or below 190 °F only

one charging pump will be allowed to

be capable of injecting into the RCS.

This restriction will continue to apply

until the RCS is vented through a

passive vent [greater than or equal to]

2.2 in 2. If this passive vent size is

established, two charging pumps and

one HPSI pump are allowed to be

capable of injecting.

The passive vent required if one or

two power operated relief valves

(PORVs) are inoperable (Technical

Specification Action Statements

(TSASs) a, b, and c) will be changed

from 2.8 in 2 or 1.4 in 2 to 2.2 in 2. A

passive vent of 1.4 in 2 is equivalent to

the vent area of one PORV. Since the

LTOP analysis assumes 2 operable

PORVs initially, and then one PORV

fails to actuate, RCS overpressure

protection will be ensured by a passive

vent of 1.4 in 2. However, a passive vent

is established by removing a pressurizer

PORV or the pressurizer manway, the

normal vent path. The value of 2.2 in 2

is the minimum size of vent that will

ensure RCS pressure remains [less than

or equal to] 300 psia, which is more

conservative than the Appendix G

limits. This vent size will also ensure

that RCS pressure does not exceed the

SDC System design pressure. In

addition, this is the size of vent that will

satisfy Technical Specification 3.4.9.1 to

allow a 50 °F/hr cooldown rate below

190 °F.

TSAS d will be added to address

excessive pumping capacity. The

required completion time of

‘‘immediate’’ reflects the importance of

this restriction, and is consistent with

current Technical Specification

requirements (Technical Specification

3.1.2.3 TSAS b and Technical

Specification 3.5.3 TSAS c) for this

situation.

These proposed changes are all more

restrictive than the previous

requirements, except for allowing 2

charging pumps and one HPSI in Mode

5 between 200 °F and 190 °F, and

requiring a vent of 2.2 in 2 instead of 2.8

in 2 when two PORVs are inoperable.

However, the proposed mass input and

venting restrictions are consistent with

the new LTOP analysis. This analysis

has demonstrated that with the

proposed restrictions the required LTOP

system will provide adequate protection

for RCS overpressurization transients.

Therefore, the proposed changes will

not result in a significant increase in the

probability or consequences of an

accident previously evaluated.

Increase in Technical Specification

Applicability. The applicability of

Technical Specification 3.4.9.3 will be

expanded to include all of Mode 5, and

Mode 6 until the reactor vessel head is

removed. The current applicability is

limited in Modes 5 and 6 to when the

RCS is not vented through a vent

[greater than or equal to] 2.8 in 2.

Expanding the applicability will ensure

an LTOP System is in place, except

when RCS pressurization is not possible

(reactor vessel head removed). This will

ensure the 10 CFR 50 Appendix G limits

are not exceeded.

The applicability of Technical

Specification 3.4.9.1 will be expanded.

The current applicability is Modes 1

through 5. However, concern for nonductile failure of the reactor vessel and

flange applies at all times, not just in

Modes 1 through 5. Therefore, the

applicability will be expanded.

Increasing the applicability of Technical

Specification 3.4.9.1 will place

additional restrictions on the plant.

However, these additional restrictions

will ensure the integrity of the RCS, in

particular the reactor pressure vessel, is

maintained. Therefore, the RCS will

continue to function as designed.

These more restrictive changes will

not result in a significant increase in the

probability or consequences of an

accident previously evaluated.

LTOP PORV Setpoint Change. The

required PORV actuation setpoint will

be reduced from [less than or equal to]

450 psig to [less than or equal to] 415

psia (400 psig). The 50 psi setpoint

reduction (pressure units have also been

changed to agree with control room

indication) will cause the PORVs to

actuate earlier during an LTOP transient

to prevent an RCS overpressurization. It

is a more restrictive change that is

consistent with the new LTOP analysis.

PORV actuation at the proposed

setpoint, in combination with the

proposed mass input restrictions, will

ensure the 10 CFR 50 Appendix G limits

are not exceeded. Therefore, the

proposed change will not result in a

significant increase in the probability or

consequences of an accident previously

evaluated.

RCP Start Criteria. The requirements

to start the first reactor coolant pump

(RCP), when RCS temperature is [less

than or equal to] 275 °F, will be

modified. The new criteria will ensure

that starting an RCP will not result in an

energy addition transient that could

exceed the capability of the steam

bubble in the pressurizer to mitigate the

event. (No credit for PORV actuation

during this energy addition transient

was assumed in the new LTOP

analysis.) This will ensure that the 10

CFR 50 Appendix G limits are not

exceeded.

The proposed RCP restrictions are

consistent with the new LTOP analysis.

This analysis has demonstrated that

with the proposed restrictions the

pressurizer will provide adequate

protection for RCS overpressurization

transients. Therefore, the proposed

changes will not result in a significant

increase in the probability or
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4317

consequences of an accident previously

evaluated.

Boron Dilution Analysis. The analysis

of the boron dilution event contained in

the Millstone Unit No. 2 FSAR [Final

Safety Analysis Report] Section 14.4.6

assumes that dilution flow rate is

limited to 88 gpm in Modes 4, 5, and 6.

Since the charging pumps are the

assumed dilution source, no more than

two charging pumps can be injecting for

this assumption to remain valid. This

results in a Technical Specification

requirement that no more than two

charging pumps can be operable when

the RCS is in Mode 4 or below (< 300

°F). This requirement will be modified

by replacing the word ‘‘operable’’ with

‘‘capable of injecting into the RCS.’’

This more accurately addresses the

boron dilution analysis restriction of

limiting the dilution flow to two

charging pumps since an inoperable

pump can still inject into the RCS. This

change is consistent with the boron

dilution accident analysis. The boron

dilution analysis further assumes that if

this dilution flow rate restriction is met,

there will be sufficient time for the

operators to recognize and terminate the

dilution before a complete loss of

shutdown margin occurs. Operator

action to restore shutdown margin by

boration is not assumed.

The proposed changes will not affect

the current Technical Specification

restriction that no more than two

charging pumps can be capable of

injecting into the RCS (operable) when

the RCS is below 300 °F. However, no

corresponding action statement

currently exists in Technical

Specification 3.1.2.4 to provide

guidance if this requirement is not met.

The addition of the proposed action

statement to Technical Specification

3.1.2.4 will require immediate action to

correct this situation. This is consistent

with other current Technical

Specification requirements (Technical

Specification 3.1.2.3 TSAS b and

Technical Specification 3.5.3 TSAS c)

that address excessive RCS makeup

capacity. Therefore, the proposed

changes will not result in a significant

increase in the probability or

consequences of an accident previously

evaluated.

RCS Pressure/Temperature and

Heatup/Cooldown Limit Changes. The

proposed changes to the heatup and

cooldown rates are a result of the new

analysis of the RCS pressure/

temperature and heatup/cooldown

limits. These changes will provide

flexibility during plant heatup and

cooldown, and especially during

equipment manipulations such as

securing RCPs, swapping shutdown

cooling (SDC) heat exchangers, and

initiating SDC.

Figure 3.4.2 will be replaced by two

curves, Figures 3.4–2a and 3.4–2b. Each

figure will contain the minimum flange

boltup temperature and the minimum

temperature for criticality. The heatup

figure (Figure 3.4–2a) will also contain

the inservice leak and hydrostatic

testing limits. The temperature change

limits will be contained in the new

Table 3.4–2, instead of in the LCO

[limiting condition for operation]. The

new limits will use cold leg temperature

instead of average temperature to

determine when to change rates. There

should be little difference between these

two temperatures, and cold leg

indication is directly available to the

control room operators.

The proposed curves and rates are

based on indicated cold leg temperature.

This parameter, which is the best

available indication of reactor vessel

downcomer temperature, will normally

be monitored by using either RCS cold

leg temperature indication or SDC

return temperature. Plant conditions

will determine which one is the

appropriate indication to use. Actual

RCS cold leg temperature will be used

if any RCP is operating or natural

circulation is occurring. Otherwise, SDC

return temperature will be used.

RCP restrictions, assumed in the

development of the heatup and

cooldown curves, will be added to the

curves. Most of the RCP restrictions

already exist either in other Technical

Specifications or in plant procedures.

Two new RCP restrictions will be added

to Technical Specifications. The

restriction of no more than three RCPs

until RCS temperature is above 500 °F

already exists in plant procedures, but

it will be added to Technical

Specifications. The restriction of no

more than two RCPs when RCS

temperature is below 200 °F already

exists in Technical Specifications

(3.4.1.4). The RCP restriction of no RCPs

below 150 °F during plant cooldown

will be added to Technical

Specifications. This restriction will have

no effect on plant operations because

RCPs will normally be secured when

cooling down below 150 °F to minimize

heat input.

The inservice leak and hydrostatic

testing temperature change limit

currently specified in Technical

Specification 3.4.9.1.c will be relocated

to Table 3.4–2. The wording will be

modified (clarification only) to specify

the limit also applies for one hour prior

to the start of inservice leak and

hydrostatic testing. This is necessary

since the development of the inservice

leak and hydrostatic testing test curve

assumes isothermal conditions. The

wording will also be modified to specify

the restrictions apply during testing

above the heatup curve instead of above

system design pressure. This type of

testing is not performed above system

design pressure.

The 50 °F/hr cooldown rate and curve

will normally be used when the RCS is

<190 °F and an RCS vent of >2.2 in2 has

been established. This curve and rate

may also be used when RCS cold leg

temperature is below 230 °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 °F), instead of the current

requirement to be <200 °F. This slight

relaxation will have no significant

impact on plant operations because

plant temperature is not normally

maintained at the mode change

temperature limit.

The required RCS pressure will be

reduced from 500 psia to 300 psia. This

is closer to the actual plant conditions

established in Mode 5. The required

RCS pressure will be reduced from 500

psig to 500 psia for the Pressurizer,

Technical Specification 3.4.9.2. The

change in units is consistent with plant

instrumentation. Establishing a lower

RCS pressure is more conservative

because it will result in less pressure

stress on either the reactor vessel or the

pressurizer.

In other than Modes 1 through 4,

immediate action will be required for

limit restoration. Violation of these

limits is typically more severe when the

RCS is cold (<200 °F), therefore an

immediate response is appropriate. A

time limit of prior to entering Mode 4

will be placed on the performance of the

engineering evaluation. This will

prevent plant startup until the

evaluation has determined that the RCS

is acceptable for continued operation.

The frequency of surveillance

requirements 4.4.9.1.c, 4.4.9.2, 4.10.3.1

will be increased from once per hour to

once per 30 minutes. This more

restrictive change will provide the plant

operators with earlier indication that a

limit may be exceeded, so that action

can be taken to prevent exceeding the

limit. The proposed changes to the RCS
4318 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

pressure/temperature limits and

temperature change rates are based on

the new analysis. This analysis uses

standard approved methods that ensure

the margins of safety required by 10 CFR

50, Appendix G are maintained. The

other changes discussed are more

restrictive enhancements to Technical

Specification requirements. Therefore,

the proposed changes will not result in

a significant increase in the probability

or consequences of an accident

previously evaluated.

Other Changes. The scope of the

action statement for Technical

Specification 3.1.2.6 will be expanded

to cover all three flowpaths identified in

Technical Specification 3.1.2.2.a. The

intent of the current wording is to

address all flowpaths. These minor

wording changes will meet this intent.

Clarification will be added to SR

4.1.2.8.d to be consistent with SR

4.1.2.7.c. The clarification will allow the

boric acid storage tank (BAST)

temperature to be verified by checking

the ambient air temperature.

A note will be added to Technical

Specification 3.4.3 TSAS a to allow the

block valve(s) to be cycled during plant

cooldown when the block valve(s)

is(are) closed due to inoperable

PORV(s). The footnote will allow the

PORV block valve(s) to be cycled during

a plant cooldown to prevent thermal

binding. This will ensure the associated

block valve(s) can be opened to allow

the PORV(s) which is(are) inoperable,

can be manually cycled if necessary.

Therefore, the PORV block valve(s) will

be able to function as designed.

The wording of Technical

Specification 3.4.3 Action Statement d

will be revised to state what action

should be performed, and to remove

specific details on how to perform the

required action. This does not change

the intent of the action statement.

Therefore, the pressurizer PORVs will

continue to function as designed.

An action statement will be added to

Technical Specification 3.4.9.3 to

provide an exception to Technical

Specification 3.0.4 requirements. This is

necessary to allow a plant cooldown to

MODE 5 if one or both PORVs are

inoperable. MODE 5 conditions may be

necessary to repair the PORV(s).

A footnote will be added to Technical

Specification 3.5.3 to allow entry into

Mode 4 without an operable high

pressure safety injection pump. This

new footnote will allow the plant to

enter Mode 4 where this specification is

applicable without any operable HPSI

pumps. However, this condition will

only be allowed for a very short time

period, one hour. The proposed change

to Technical Specification 3.4.9.3 will

allow a HPSI pump to be operable above

190 °F. However, the 10 °F range before

Mode 4 is reached may not allow

sufficient time to ensure a HPSI pump

is operable. Adding this note will

provide the operating crew sufficient

time to make an orderly transition into

Mode 4. This condition will only be

allowed for one hour, which is the same

time allowed by the first part of TSAS

a for an inoperable HPSI pump.

The LTOP requirements currently

contained in Technical Specifications

3.1.2.3 and 3.5.3 will be relocated to the

LTOP Technical Specification 3.4.9.3.

Relocating requirements within

Technical Specifications will not

change the technical content of the

requirement.

Various redundant or outdated

Technical Specification requirements

will be eliminated and references will

be adjusted to reflect the proposed

changes. Removal of redundant or

outdated requirements from Technical

Specifications and adjustments to

references to other requirements will

not impact any technical requirements.

Minor wording changes have been

made to many of the Technical

Specifications contained in this license

amendment request. These changes do

not change any technical aspect of the

Technical Specification affected. They

are editorial changes only.

The proposed changes do not alter the

way any structure, system, or

component functions. There will be no

effect on equipment important to safety.

Therefore, the proposed changes will

not result in a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes have no effect

on any of the design basis accidents

previously evaluated. Therefore, the

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 will 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 changes will modify the

LTOP requirements, RCS pressure/

temperature limits, and the RCS heatup

and cooldown limits. The majority of

the proposed changes are being made as

a result of the new pressure/temperature

and LTOP analyses performed. The new

pressure/temperature curves and heatup

and cooldown rates were developed in

accordance with the requirements and

methods described in 10 CFR 50

Appendix G and are consistent with the

criteria contained in the Standard

Review Plan Section 5.3.2. The new

LTOP mass input and RCP starting

restrictions and LTOP PORV setpoints

are consistent with the criteria

contained in the Standard Review Plan

Section 5.2.2. Additional changes have

been proposed to correct various items

identified during the review of the

Millstone Unit No. 2 Technical

Specifications. The proposed changes

do not change the requirements to

maintain RCS pressure and temperature

within the requirements defined in

Technical Specifications. This will

ensure the integrity of the reactor vessel

is maintained during all aspects of plant

operation. Therefore, there is no

significant effect on the probability or

consequences of any accident

previously evaluated and no significant

impact on offsite doses associated with

previously evaluated accidents. This

License Amendment Request does not

result in a reduction of the margin of

safety as defined in the Bases for the

Technical Specifications addressed by

the proposed changes.

The NRC has provided guidance

concerning the application of standards

in 10 CFR 50.92 by providing certain

examples (March 6, 1986, 51 FR 7751)

of amendments that are considered not

likely to involve an SHC [significant

hazards consideration]. The changes

proposed herein to correct terminology,

numbering, references, and relocating

requirements within Technical

Specifications are enveloped by

example (i), a purely administrative

change to Technical Specifications. The

more restrictive changes proposed

herein that are based on the new

analyses performed and the more

restrictive enhancements are enveloped

by example (ii), a change that

constitutes an additional limitation,

restriction, or control not presently

included in Technical Specifications.

All other changes proposed herein are

not enveloped by a specific example.

As described above, this License

Amendment Request does not impact

the probability of an accident previously

evaluated, does not involve a significant
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4319

increase in the consequences of an

accident previously evaluated, does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated, and does

not result in a significant reduction in

a margin of safety. Therefore, NNECO

[Northeast Nuclear Energy Company]

has concluded that the proposed

changes do not involve an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

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:

December 8, 1997.

Description of amendment request:

The proposed amendment would

change the Technical Specifications

(TSs) to resolve several compliance

issues. The proposed changes would (1)

correct the wording and the formula in

TS Definition 1.18 ‘‘Azimuthal Power

Tilt—Tq;’’ (2) correct the wording in TS

4.1.1.1.2 ‘‘Reactivity Control Systems

Shutdown Margin—Tavg [less than or

equal to] 200 °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 <300 °F;’’ (6)

correct the wording, frequency, and

reference number for the surveillance

requirements in TS 3.6.3.2

‘‘Containment Systems Containment

Ventilation System;’’ (7) correct the

nomenclature used for the A.C. busses

in TSs 3.8.2.1 and 3.8.2.1A ‘‘Onsite

Power Distribution Systems A.C.

Distribution—Operating;’’ (8) correct TS

Bases by modifing the applicable

sections to reflect the proposed changes;

(9) delete the word ‘‘original’’ from the

statement ‘‘original design provision’’ in

Design Features Section—TSs 5.1.3

‘‘Flood Control,’’ 5.2.3 ‘‘Penetrations,’’

5.3.2 ‘‘Control Element Assemblies,’’

and 5.7.1 ‘‘Seismic Classification;’’ and

(10) delete Design Features Section—TS

5.9 ‘‘Shoreline Protection.’’

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 in the wording

and associated formula of Technical

Specifications Definition 1.18 will

ensure the calculated value of

Azimuthal Power Tilt (Tq) 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
4320 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

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
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 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

margin of safety?

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. 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). The proposed change

involves no physical modification of the

plant, or alterations to any accident or

transient analysis. [* * *] Therefore,

the change does not involve any

significant reduction in a margin of

safety.

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, 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 involve any 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: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request:

December 19, 1997.

Description of amendment request:

The proposed amendment would revise

the Hope Creek Generating Station

(HCGS) Technical Specifications (TS) to

incorporate changes that reflect the

completion of the Salt Drift Monitoring

Program.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
4322 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The changes, which update the

Terrestrial Ecology Monitoring Program

status, are administrative in nature and

in no way affect the initial conditions,

assumptions, or conclusions of the Hope

Creek Generating Station accident

analyses. In addition, the proposed

changes would not affect the operation

or performance of any equipment

assumed in the accident analyses. Based

on the above information, we conclude

that the proposed changes would not

significantly increase the probability or

consequences of an accident previously

evaluated.

2. The proposed changes do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

As previously stated, the proposed

changes are administrative in nature

and in no way impact or alter the

configuration or operation of the

facilities and create no new modes of

operation. PSE&G therefore concludes

that the proposed changes would not

create the possibility of a new or

different kind of accident.

3. The proposed changes do not

involve a significant reduction in a

margin of safety.

The changes are administrative in

nature and in no way affect plant or

equipment operation or the accident

analysis. PSE&G therefore concludes

that the proposed changes would not

result in 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: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request:

November 14, 1997.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications

(TSs) to provide surveillance

requirements for the service water

accumulator vessels. Specifically,

surveillance requirements are provided

for vessel level, pressure and

temperature, and discharge valve

response time. The surveillance

requirements are included in TS 3/

4.6.1.1 and 3/4.6.2.3, and the applicable

Bases sections are expanded to provide

supporting information.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes provide

surveillance requirements for the

Service Water [SW] accumulator tank

level, pressure and temperature

parameters and the discharge valve

response time test. Supporting

information is included in the Bases

section of the applicable technical

specifications. The SW accumulator

tank and discharge valve design has

been reviewed and approved by the

NRC staff as documented in NRC Safety

Evaluation Report (SER) dated June 19,

1997. The proposed surveillance

requirements do not alter the design as

reviewed by the NRC staff. The addition

of tank parameter surveillance

requirements to the technical

specifications does not alter the

physical plant arrangement or the

installed monitoring instrumentation.

The proposed addition of tank discharge

valve response time surveillance

requirements to the technical

specifications does not alter the method

of performing these surveillance

requirements.

Therefore the proposed changes do

not increase the probability of an

accident. The surveillance requirements

provide additional controls for ensuring

the SW accumulator tank and discharge

valves will be maintained within the

design parameters assumed in the safety

analysis. This provides added assurance

that the accumulator tanks and

discharge valves will be capable of

performing their required design

function during accident conditions.

There is no change to the performance

requirements of these components in

preventing two phase flow conditions

and water column separation

waterhammer vulnerabilities identified

in GL [Generic Letter] 96–06. Therefore,

the proposed changes do not involve an

increase in the consequences of an

accident previously evaluated.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes provide

surveillance requirements for Service

Water Accumulator tank level, pressure

and temperature and discharge valve

time response. Supporting information

is included in the Bases section of the

applicable technical specifications. The

SW accumulator tank and discharge

valve design has been reviewed and

approved by the NRC staff as

documented in NRC Safety Evaluation

Report (SER) dated June 19, 1997. The

proposed surveillance requirements do

not alter the plant configuration.

Installed instrumentation will be used

to accomplish the tank surveillance

requirements. The current plant

installation also provides for completion

of the discharge valve response time

surveillance utilizing test equipment in

accordance with plant procedures and

configurations. Therefore the

performance of these surveillance

requirements does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The Service Water Accumulator

Vessels and discharge valves were

installed to address the Generic Letter

96–06 issues of column separation

waterhammer and two phase flow in the

containment fan coil unit (CFCU) piping

during an accident involving loss of

offsite power. This design has been

reviewed and approved by the NRC staff

as documented in NRC Safety

Evaluation Report (SER) dated June 19,

1997. The proposed surveillance

requirements do not alter the design as

reviewed by the NRC staff. By providing

added assurance that these components

are capable of performing their specified

safety function as assumed in the safety

analysis, the additional surveillance

requirements assure system operability

to further minimize the possibility of

waterhammer and two phase flow in the

CFCU piping during accident

conditions. The proposal therefore

minimizes the possibility of a new or

different kind of accident from those

previously evaluated accidents.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

The additional surveillances provide

added assurance that the margin of

safety assumed in the containment

integrity and containment cooling

technical specification will be
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4323

maintained. The additional surveillance

requirements further ensure that in the

event the SW accumulator vessels are

out of specification or the discharge

valves do not meet their response time

requirements, corrective actions will be

completed in accordance with the

existing containment integrity technical

specification allowed outage time to

restore containment integrity. The

surveillance requirements further

ensure that in the event the SW

accumulator vessel or discharge valves

do not meet these requirements,

corrective actions will be completed in

accordance with the containment

cooling technical specification allowed

outage time to restore the full

complement of containment fan coil

units to operability. Since the proposal

maintains the margin of safety provided

in the containment integrity and

containment cooling technical

specification, there is no 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: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

Southern California Edison Company,

et al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of amendment requests:

September 16, 1997.

Description of amendment requests:

The licensee proposes to revise

Technical Specification (TS) 3.4.13,

‘‘RCS Operational Leakage,’’ TS

5.5.2.11, ‘‘Steam Generator (SG) Tube

Surveillance Program,’’ and TS 5.7.2,

‘‘Special Reports.’’ The proposed change

is to allow steam generator tube repair

using ASEA Brown Boveri/Combustion

Engineering (ABB/CE) leak tight

sleeving as an alternative steam

generator tube repair to 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. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The supporting technical evaluation

and safety evaluation for the ASEA

Brown Boveri/Combustion Engineering

(ABB/CE) leak tight sleeves demonstrate

that the sleeve configuration will

provide steam generator (SG) tube

structural and leakage integrity under

normal operating and accident

conditions. The sleeve configurations

have been designed and analyzed in

accordance with the requirements of the

ASME Code. Mechanical testing has

shown that the sleeve and sleeve joints

provide margin above acceptance limits.

Ultrasonic Testing (UT) is used to verify

the leak tightness of the weld above the

tubesheet. Testing has demonstrated the

leak tightness of the hardroll joint due

to the reinforcing effect of the tubesheet.

Tests have demonstrated that tube

collapse will not occur due to

postulated Loss of Coolant Accident

(LOCA) loadings.

A new, more conservative, Technical

Specification (TS) SG tube leakage rate

requirement is introduced by this

change. Accident analysis assumptions

remain unchanged in the event that

significant leakage does occur from the

sleeve joint or that the sleeve assembly

ruptures. Any leakage through the

sleeve assembly is fully bounded by the

existing SG tube rupture analysis

included in the San Onofre Nuclear

Generating Station (SONGS) Updated

Final Safety Analysis Report. Reactor

coolant flow reduction from sleeving is

addressed by a ratio of number of tubes

sleeved to equal a plugged tube. The

proposed sleeving repair process does

not adversely impact any other

previously evaluated design basis

accidents.

Therefore, proposed changes do not

involve a significant increase in the

probability or consequences of an

accident.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

Installation of the sleeves does not

introduce any significant changes to the

plant design basis. The use of a sleeve

to span the area of degradation of the SG

tube restores the structural and leakage

integrity of the tubing to meet the

original design bases. Stress and fatigue

analysis of the sleeve assembly shows

that the requirements of the ASME Code

are met. Mechanical testing has

demonstrated that margin exists above

the design criteria. Any hypothetical

accident as a result of any degradation

in the sleeved tube would be bounded

by the existing tube rupture accident

analysis.

Therefore, the operation of the facility

in accordance with proposed changes

does 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 use of sleeves to repair degraded

SG tubing has been demonstrated to

maintain the integrity of the tube bundle

commensurate with the requirements of

the ASME Code and draft Regulatory

Guide (RG) 1.121 and to maintain the

primary to secondary pressure boundary

under normal and postulated accident

conditions. The safety factors used in

the verification of the strength of the

sleeve assembly are consistent with the

safety factors in the ASME Boiler and

Pressure Vessel Code used in SG design.

The operational and faulted condition

stresses and cumulative usage factors

are bounded by the ASME Code

requirements. The sleeve assembly has

been verified by testing to prevent both

tube pullout and significant leakage

during normal and postulated accident

conditions. A test program was

conducted to ensure the lower

hardrolled joint design was leak tight

and capable of withstanding the design

loads. The primary coolant pressure

boundary of the sleeve assembly will be

periodically inspected by NonDestructive Examination to identify

sleeve degradation due to operation.

Installation of the sleeves will

decrease the number of tubes which

must be taken out of service due to

plugging. There is a small amount of

primary coolant flow reduction due to

the sleeve for which the equivalent

sleeve to plug ratio is assigned based on

sleeve length. The ratio is used to assess

the final equivalent plugging percentage

as an input to other safety analyses. The

sleeve maintains the design basis

requirements for the SG tubing.

Therefore, operation of the facility

with the proposed changes 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 requests

involve no significant hazards

consideration.

Local Public Document Room

location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison
4324 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

Company, P. O. Box 800, Rosemead,

California 91770.

NRC Project Director: William H.

Bateman.

Southern California Edison Company,

et al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of amendment requests: October

17, 1997.

Description of amendment requests:

The licensee proposes to amend the

licenses for SONGS Units 2 and 3 to

revise the Final Safety Analysis Report

(FSAR) to permit digital radiation

monitor installation for both trains

supplying the containment purge

isolation signal, and permit digital

radiation monitor installation for both

trains supplying the control room

isolation signal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed change is required to

permit using digital radiation monitors

as input to both trains of the Control

Room Isolation Signal (CRIS), and to

both trains of the Containment Purge

Isolation Signal (CPIS). These changes

will allow replacement of the remaining

safety related obsolete radiation monitor

equipment to address spare parts and

equipment availability issues. The new

containment airborne radiation digital

monitor will have the same basic

architecture as the existing analog

system, and serves to perform the same

function. In addition, the digital

radiation monitors are expected to be

more reliable than the existing

equipment which is of an analog design.

Furthermore, defense-in-depth

equipment is available that either

provides, or allows for, actions to

mitigate the release of offsite and

Control Room doses to within existing

licensing limits based on realistic event

input assumptions. Analyses show that

if ‘‘realistic’’ input assumptions are

utilized and reasonable operator actions

are allowed, then acceptable dose

consequences result both to the general

public offsite, and to the Control Room

operators.

Therefore, the proposed change will

not involve a significant increase in the

probability or consequences of any

accident previously evaluated.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed change will permit

upgrading the existing analog radiation

monitors with upgraded digital

radiation monitors. Replacement of an

analog system to a predominantly

digital system, uses software algorithms

to perform the required functions. A

satisfactory software verification and

validation (V&V) report, including

continued software change control

procedures, provides assurance that a

software common mode failure is not

likely.

In addition, the design, installation,

testing, maintenance, and operation of

the affected equipment will assure that

no new or different kinds of accidents

will be created. The ESFAS radiation

monitors involved are portions of

systems that respond to accidents. They

can not, by their actions or inactions,

create a new or different accident from

any accident previously evaluated.

Therefore, the proposed change will

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 CRIS and CPIS Radiation Monitor

Systems provide an accident mitigation

function for offsite doses (10 CFR 100)

and Control Room doses (10 CFR 50

Appendix A, General Design Criteria

19). A change in the margin of safety is

introduced due to the possibility of a

software common mode failure in

redundant equipment simultaneously

affecting equipment performing a

different function.

This change is not a significant

reduction in the margin of safety,

however, due to the following:

(1) A probabilistic risk analysis has

determined that the availability of the

affected radiation monitors, including

software, should be better than the

existing equipment based on industry

data to date,

(2) The software V&V and

preoperational testing to be performed

will provide assurance of system

operation, and

(3) The combined occurrence of a

software common mode failure that

simultaneously causes failure of all

available ESFAS radiation monitors

concurrent with a design bases accident

is very unlikely.

In the unlikely event of a software

common mode failure that causes all

ESFAS radiation monitors to be

inoperable concurrent with a design

bases accident, analyses show that if

‘‘realistic’’ input assumptions are

utilized and reasonable operator actions

are allowed, then acceptable dose

consequences result both to the general

public offsite, and to the Control Room

operators.

Therefore, the proposed change will

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770.

NRC Project Director: William H.

Bateman.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket

Nos. 50–321 and 50–366, Edwin I.

Hatch Nuclear Plant, Units 1 and 2,

Appling County, Georgia

Date of amendment request:

December 18, 1997.

Description of amendment request:

The proposed amendments would

modify or delete obsolete conditions

from the Unit 1 and Unit 2 Operating

Licenses. The changes are editorial or

administrative in nature.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes either remove

or modify provisions in the Plant Hatch

Unit 1 and Unit 2 Operating Licenses

that have been completed or are

otherwise obsolete. Certain Surveillance

Requirements (SRs) that were either

added or modified at the time of

Improved Technical Specifications (ITS)

implementation were listed in the

Operating Licenses with a schedule for

performance. With the exception of Unit

1 SR 3.8.1.18, all SRs are deleted from

the Operating Licenses, because they

have since been performed according to

schedule, and will henceforth be
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4325

performed in accordance with the

Technical Specifications.

A requirement for submittal of the

Unit 1 inservice inspection plan for the

recirculation and residual heat removal

systems’ piping is deleted due to

completion of the activity.

Two exemptions granted at Unit 2

startup are deleted due to completion of

the required activities associated with

the exemptions. These were seismic

qualification demonstration for the Unit

2 reactor protection system power

supply and completion of the long-term

BWR [boiling water reactor] Owner’s

Group Mark I containment program.

A requirement to conduct the Unit 2

Initial Test Program according to the

requirements in Chapter 14 of the Final

Safety Analysis Report without major

changes is deleted due to completion of

the activity. A condition relating to

environmental protection is deleted

from the Unit 2 Operating License, since

it was superseded by the Environmental

Protection Plan (Nonradiological),

Appendix B to the Operating Licenses.

Attachment 2, Items To Be Completed

Prior To Opening Main Steam Isolation

Valves, is deleted due to completion of

the activities.

The proposed changes discussed

above are strictly administrative/

editorial and do not affect the operation

or function of any plant system,

component, or structure. Therefore, the

proposed changes do not increase the

probability of occurrence or the

consequences of a previously evaluated

accident.

2. The proposed changes do not create

the possibility of a new and different

type of accident from any previously

evaluated.

The proposed administrative/editorial

changes do not alter the operation of

any plant system or equipment and do

not introduce a new mode of operation.

Thus, the proposed changes cannot

create a new accident initiating

mechanism. Therefore, the proposed

changes do not create the possibility of

a new and different type of accident

from any previously evaluated.

3. The proposed changes do not

involve a significant reduction in the

margin of safety.

Since the proposed changes are

strictly administrative/editorial and do

not involve any physical or procedural

changes to the plant, the margin of

safety, as defined in the bases for any

Technical Specification is not 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: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC.

NRC Project Director: Herbert N.

Berkow.

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:

December 17, 1997.

Description of amendment request:

The proposed amendment would extend

the surveillance interval of the

containment spray system nozzle air

flow test from five years to ten years.

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. Operation of the facility in

accordance with the proposed

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not result

in any hardware changes. The

Containment Spray system trains or

nozzles are not assumed to be the

initiators of any analyzed events.

Extending the surveillance interval for

performing the Containment Spray

system nozzle air flow test from five to

ten years does not represent a

significant increase in the probability of

an accident. The Containment Spray

system nozzles are not precursors to any

accident analyses.

The Containment Spray system trains

and nozzles function to mitigate the

consequences of an analyzed event by

providing spray flow to containment

during an accident. The proposed

change still provides assurance that the

Containment Spray system nozzles will

be maintained operable due to the

passive nature of the design, the

materials of construction, and the lowstress non-wetted environment. The

extension of the surveillance interval

does not significantly increase the

probability or consequences of an

accident since the nozzle will still be

OPERABLE between surveillance tests.

B. Operation of the facility in

accordance with the proposed

amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not

necessitate a physical alteration of the

plant or changes in parameters

governing normal plant operation. No

new or different types of equipment will

be installed. The proposed change will

still ensure Containment Spray system

nozzle OPERABILITY is adequately

maintained.

C. Operation of the facility in

accordance with the proposed

amendment does not involve a

significant reduction in a margin of

safety.

The increased interval between the

Containment Spray system nozzle air

flow test is acceptable due to the passive

design of the nozzles and industry

operating experience as detailed in

NURG–1366. The increased interval is

considered acceptable for maintaining

nozzle OPERABILITY. The Containment

Spray system, including the nozzles,

will continue to provide their required

safety function with the increase from

five to ten years between inspections.

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, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036–5869.

NRC Project Director: John Hannon.

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:

December 31, 1997.

Description of amendment request:

The proposed amendment would revise

Technical Specifications 2.1 (Safety

Limits), 2.2 (Limiting Safety System

Settings), and 3/4.2.5 (Departure from

Nucleate Boiling Parameters) by

including alternate operating criteria to

allow continued plant operation with a

reduced measured reactor coolant

system flow rate, if necessary.

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
4326 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

consideration, which is presented

below:

(1) Does the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

evaluated?

The affected Reactor Protection

System functions will continue to

provide their current safety function

under alternate operating criteria for

reduced measured Reactor Coolant

System flow conditions. The OT DeltaT [Overtemperature Delta-T], OP DeltaT [Overpower Delta-T], and f(Delta-I) [a

function of the indicated difference

between top and bottom detectors of the

power-range neutron ion chambers]

safety-analysis reactor trip setpoints

have been recalculated to appropriately

reflect the reduced flow conditions. In

doing so, the difference, or margins,

between the nominal and maximum

values of the reference trip setpoints

(i.e., K1, and K4 for the OT Delta-T and

the OP Delta-T setpoints, respectively)

have been maintained so that the Total

Allowance remains unchanged and,

therefore, the instrument accuracy

uncertainties are unaffected.

Furthermore, implementation of the

provisions for reduced measure Reactor

Coolant System flow under alternate

operating criteria for the South Texas

Project Technical Specifications does

not increase the probability or

consequences of an accident previously

evaluated in the UFSAR [Updated Final

Safety Analysis Report]. This change

cannot directly initiate an accident. The

consequences of accidents previously

evaluated in the UFSAR are unaffected

by this proposed change because no

change to any equipment response or

accident mitigation scenario has

resulted. There are no additional

challenges to fission product barrier

integrity. Therefore, the probability of

an accident previously evaluated has

not been increased.

(2) Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any accident previously evaluated?

No new failure mechanisms or

accident scenarios or limiting single

failures are introduced as a result of this

proposed change. Operation of the plant

will be consistent with that previously

modeled. All of the accident analyses

previously evaluated in the UFSAR for

South Texas Project Units 1 and 2 have

been evaluated to support alternate

operating condition with a 3 percent

reduction in the minimum measured

Reactor Coolant System flow. The new

nominal Reactor Coolant System

operating conditions supported by these

evaluations have been determined.

Revised Core Thermal Safety Limits

have been established and will be

incorporated into the Technical

Specifications for the 3 percent Reactor

Coolant System measured flow

reduction; and, the OT Delta-T and OP

Delta-T setpoints are re-calculated based

on the new Safety Analysis Limits,

appropriate for the reduced flow

operation. These reactor protection

system functions affected by the change

in operating conditions will, therefore,

continue to provide an appropriate

response equivalent to current safety

analysis modeling. The proposed

Technical Specification amendment

does not challenge the performance or

integrity of safety-related systems. The

possibility of a new or different kind of

accident, therefore, is not created.

(3) Does the proposed amendment

involve a significant reduction in a

margin of safety?

The modification will have no effect

on the availability, operability, or

performance of the South Texas Project

safety-related systems and components.

This is based on: the evaluation

performed of all accidents previously

evaluated in the UFSAR for operation of

South Texas Project Units 1 and 2 at

reduced Reactor Coolant System flow

conditions; establishment of revised

Core Thermal Safety Limits that are

reflected in the proposed Technical

Specification applicable for the 3

percent Reactor Coolant System flow

reduction; and, the appropriately recalculated OT Delta-T and OP Delta-T

setpoints, also applicable for these

reduced flow conditions. Allowing

provision for these alternate operating

criteria does not prevent inspections or

surveillance required by the Technical

Specifications. The margin of safety

associated with the acceptance criteria

for any accident is unchanged, and

therefore, the proposed modification

will not reduce the margin of safety as

defined in the Bases of the South Texas

Project Technical Specifications.

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, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036–5869.

NRC Project Director: John N.

Hannon.

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of amendment request:

December 23, 1997.

Description of amendment request:

The license amendment request

proposes changes to technical

specification surveillances to remove

the requirements related to accelerated

testing of the standby emergency diesel

generators, consistent with the

recommendations in NRC Generic Letter

94–01, ‘‘Removal of Accelerated Testing

and Special Reporting Requirements for

Emergency Diesel Generators.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes do not

significantly increase the probability of

occurrence of a previously evaluated

accident because the standby diesel

generators (including the High Pressure

Core Spray [HPCS] diesel generator) are

not initiators of previously evaluated

accidents. The standby diesel generators

mitigate the consequences of previously

evaluated accidents involving a loss of

offsite power. The Perry Nuclear Power

Plant (PNPP) program developed to

meet the Maintenance Rule (10 CFR

50.65) will continue to ensure the diesel

generators perform their function when

called upon. The change to the

surveillance frequency does not affect

the design of the diesel generators, the

operational characteristics of the diesel

generators, the interfaces between the

diesel generators and other plant

systems, the function, or the reliability

of the diesel generators. Thus, the diesel

generators will be capable of performing

their accident mitigation function, there

is no impact to the radiological

consequences of any accident analysis,

and the probability and consequences of

previously evaluated accidents are not

increased by this activity.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed activity involves a

change to the frequency for specific

technical specification surveillance

requirements. No physical or
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4327

operational changes to the diesel

generators or supporting systems are

made by this activity. Since the

proposed changes do not involve a

change to the plant design or operation

and thus no new system interactions are

created by this change, these changes do

not produce any parameters or

conditions that could contribute to the

initiation of accidents different from

those already evaluated in the Updated

Safety Analysis Report. The proposed

changes only address the methods used

to ensure diesel generator reliability.

Thus, the proposed amendment does

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 changes involve the

methods used to ensure diesel generator

performance and reliability. No changes,

other than to frequency, are made to

Technical Specification Surveillance

Requirements 3.8.1.2 and 3.8.1.3. The

NRC, in Generic Letter 94–01, has

acknowledged the acceptability of the

use of the Maintenance Rule program

for the diesel generators to ensure diesel

generator performance in lieu of

accelerated testing. These proposed

changes do not involve a change to the

plant design or operation, and thus do

not affect the design of the diesel

generators, the operational

characteristics of the diesel generator,

the interfaces between the diesel

generators and other plant systems, or

the function or reliability of the diesel

generators. Because the diesel generator

performance and reliability will

continue to be ensured by the diesel

generator program to meet the

Maintenance Rule, the proposed

changes do not result in a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Richard P.

Savio.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of amendment request:

December 23, 1997.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Section 4.4.5, ‘‘Reactor Coolant

System—Steam Generators—

Surveillance Requirements (SRs).’’ SR

4.4.5.8 would be modified to provide

flexibility in the scheduling of steam

generator inspections during refueling

outages.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The Davis-Besse Nuclear Power

Station has reviewed the proposed

changes and determined that a

significant hazards consideration does

not exist because operation of the DavisBesse Nuclear Power Station, Unit No.

1, in accordance with these changes

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.

No previously analyzed accident

scenario is changed, and initiating

conditions and assumptions remain as

previously analyzed. The proposed

change to Technical Specification (TS)

Surveillance Requirement (SR) 4.4.5.8,

to allow performance of required visual

inspections of the secured internal

auxiliary feedwater header, header to

shroud attachment welds, and the

external header thermal sleeves during

the third period of the ten-year Inservice

Inspection Interval, does not affect any

Updated Safety Analysis Report (USAR)

accident initiators. These inspections

will continue to take place at a

prescribed time interval scheduled

similar to American Society of

Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code Section XI

components. Therefore, it can be

concluded that the proposed change

does 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 change does not affect

accident conditions or assumptions

used in evaluating the radiological

consequences of an accident. The

proposed change does 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 change does not alter the

way the plant is operated, and no new

or different failure modes have been

defined for any plant system or

component important to safety, nor has

any limiting single failure been

identified as a result of the proposed

changes.

These inspections were established to

ensure that there are no new failure

mechanisms resulting from these

components. These inspections will

continue to take place in the third

period of each inservice inspection

interval. No new or different types of

failures or accident initiators are

introduced by the proposed changes.

3. Not involve a significant reduction

in a margin of safety because visual

inspections will be performed on a

prescribed frequency that is consistent

with the schedules established for

ASME Code components in accordance

with ASME Code Section XI.

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.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of amendment request:

December 23, 1997.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

1.0, ‘‘Definitions,’’ to clarify the

meaning of core alteration; would

relocate TS Section 3/4.9.5, ‘‘Refueling

Operations—Communications,’’ and the

associated bases to the Technical

Requirements Manual; and would add

TS Section 3.0.6 and the associated

bases to address the return to service of

inoperable equipment.

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
4328 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

consideration, which is presented

below:

The Davis-Besse Nuclear Power

Station (DBNPS) has reviewed the

proposed changes and determined that

a significant hazards consideration does

not exist because operation of the DavisBesse Nuclear Power Station, Unit

Number 1, in accordance with these

changes would:

1a. Not involve a significant increase

in the probability of an accident

previously evaluated because the

probability of previously analyzed

accidents is not affected by the criteria

in the core alteration definition

(Technical Specification (TS) 1.12). Nor

do these changes, the proposed

relocation of the refueling

communications TS 3/4.9.5 and Bases to

the DBNPS Updated Safety Analysis

Report (USAR) Technical Requirements

Manual (TRM), or the proposed addition

of new TS 3.0.6 and Bases regarding

return to service of inoperable

equipment, affect any accident initiator,

or assumption made in any safety

analysis. The proposed changes are

administrative in nature and are

consistent with NUREG–1430, Revision

1, ‘‘Standard Technical Specifications,

Babcock and Wilcox Plants,’’ dated

April 1995, as modified by a pending

NUREG–1430 change approved by the

NRC, Technical Specification Task

Force (TSTF) Standard Technical

Specification Change Traveler Number

165.

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. The

proposed changes do not significantly

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 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.

Attorney for licensee: Jack Newman,

Al Gutterman, Morgan, Lewis &

Brockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Acting Project Director: Richard

P. Savio.

Yankee Atomic Electric Company,

Docket No. 50–029, Yankee Nuclear

Power Station, Franklin County,

Massachusetts

Date of amendment request:

December 18, 1997.

Description of amendment request: By

letter dated May 15, 1997, the licensee

submitted a License Termination Plan.

The NRC previously published a notice

dated August 14, 1997, in the Federal

Register (62 FR 43559) advising of

receipt of the Plan. The proposed

request is for a license amendment

approving the Plan for the 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:

The proposed change will not:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated. Accident

analyses are included in the approved

Decommissioning Plan and

incorporated into the FSAR. All

decommissioning and fuel storage

activities described in the License

Termination Plan are consistent with

those in the approved Decommissioning

Plan. No systems, structures, or

components that could initiate or be

required to mitigate the consequences of

an accident are affected by the proposed

change in any way not previously

evaluated in the approved

Decommissioning Plan. Therefore, the

proposed change is administrative in

nature and does not involve an 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. Accident

analyses are included in the approved

Decommissioning Plan and are

incorporated into the FSAR. All

decommissioning and fuel storage

activities described in the License

Termination Plan are consistent with

those in the approved Decommissioning

Plan. The proposed change does not

affect plant systems, structures, or

components in any way not previously

evaluated in the approved

Decommissioning Plan, and no new or

different failure modes will be created.

Therefore, the proposed change is

administrative in nature and does 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. Approval of the

License Termination Plan by license

amendment is administrative in nature

since all decommissioning and fuel

storage activities described in the

License Termination Plan are consistent

with those in the approved

Decommissioning Plan. Therefore, the

proposed change does not involve a

significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Greenfield Community

College, 1 College Drive, Greenfield,

Massachusetts 01301.

Attorney for licensee: Thomas Dignan,

Esquire, Ropes and Gray, One

International Place, Boston,

Massachusetts 02110–2624.

NRC Project Director: Seymour H.

Weiss.

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
Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices 4329

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.

Baltimore Gas and Electric Company,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

October 2, 1997.

Brief description of amendments: The

amendment changes the Calvert Cliffs

Unit 1 Technical Specification

Requirements 4.8.1.1.2.a.5, 4.8.1.1.2.d.4,

and 4.8.1.1.2.d.5. Baltimore Gas and

Electric Company is planning to modify

existing 1B emergency diesel generator

(EDG) to increase its rated continuous

capacity from 2700 kW to 3000 kW by

increasing the mechanical capacity of

the engine. The change revises the

above surveillance requirements to

reflect the new electrical capacity of 1B

EDG.

Date of issuance: January 5, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 224.

Facility Operating License Nos. DPR–

53 and DPR–69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 5, 1997 (62 FR

59913).

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated January 5, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request:

September 25, 1997.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by modifying the

Limiting Condition for Operation (LCO)

3.6.1.2 (Containment Leakage), the

associated action, and Surveillance

Requirement (SR) 4.6.1.2 for Waterford

Steam Electric Station, Unit 3

(Waterford 3). The air lock door seal

leakage rate acceptance criteria in TS

6.15 is being changed from 0.01La to

0.005La. TS 6.15 is also being modified

to make the terms used in the

Containment Leakage Rate Testing

Program consistent with terms used in

the TS.

Date of issuance: January 15, 1998.

Effective date: January 15, 1998.

Amendment No.: 138.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54872).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 15,

1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

October 10, 1997.

Brief description of amendment: The

amendment revises the Oyster Creek

Nuclear Generating Station (OCNGS)

operating license and technical

specifications to reflect the registered

trade name of ‘‘GPU Energy’’ under

which the owner of OCNGS now does

business and to reflect the change of the

legal name of the operator of OCNGS

from GPU Nuclear Corporation to GPU

Nuclear, Inc. In addition, two minor

editorial corrections associated with the

name change are included in the

amendment.

Date of issuance: January 14, 1998.

Effective date: As of the date of

issuance, with full implementation

within 30 days.

Amendment No.: 194.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 5, 1997 (62 FR

59915). The Commission’s related

evaluation of this amendment is

contained in a Safety Evaluation dated

January 14, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

October 8, 1997, and October 21, 1997.

Brief description of amendments: The

amendments increase both the

minimum required ice mass per ice

basket and the total minimum required

ice mass in the ice condenser, and

change the bases for the technical

specifications.

Date of issuance: January 2, 1998.

Effective date: January 2, 1998, with

full implementation within 45 days.

Amendment Nos.: 220 and 204.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54863).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, MI 49085

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

October 15, 1997.

Brief description of amendment:

Technical Specification Surveillances

4.1.2.3.1, 4.1.2.4.1, 4.5.2, 4.6.2.1, and

4.6.2.2 require the recirculation spray,

quench spray, residual heat removal,

centrifugal charging, and safety

injection pumps to be tested on a

periodic basis and after modifications

that alter subsystem flow characteristics.

The amendment replaces the specific

surveillance pump pressure with a

statement that the test be conducted in

accordance with Specification 4.0.5,

Inservice Testing Program. The
4330 Federal Register / Vol. 63, No. 18 /Wednesday, January 28, 1998 / Notices

amendment also decreases the required

individual safety injection and

centrifugal charging pump injection line

flow rates, increases the allowed

individual safety injection pump runout

flow rate, and makes editorial changes

to the surveillances.

Date of issuance: December 24, 1997.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 155.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 5, 1997 (62 FR

59918).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 24,

1997.

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 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

November 4, 1997.

Brief description of amendments:

These amendments revise Technical

Specification 3/4.8.1 on the emergency

diesel generators to (1) delete the 18-

month surveillance requirements

4.8.1.1.2.d.1 and (2) eliminate the

accelerated testing requirement of Table

4.8–1.

Date of issuance: January 8, 1998.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment Nos.: 203 and 185.

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 3, 1997 (62 FR

63982).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Tennessee Valley Authority, Docket No.

50–327, Sequoyah Nuclear Plant, Unit

1, Hamilton County, Tennessee

Date of application for amendments:

November 21, 1997 (TS 97–05).

Brief description of amendments: The

amendments change the Technical

Specifications (TS) to allow a one-time

provision for testing power-operated

relief valves in Mode 5.

Date of issuance: January 13, 1998.

Effective date: January 13, 1998.

Amendment No.: 230.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

TS.

Date of initial notice in Federal

Register: December 1, 1997 (62 FR

63565).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 13,

1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Dated at Rockville, Maryland, this 21st day

of January 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–1904 Filed 1–27–98; 8:45 am]

BILLING CODE 7590–01–P

PEACE CORPS

Information Collection Requests Under

OMB Review

ACTION: Notice of public use form

review request to the Office of

Management and Budget.

SUMMARY: Pursuant to the Paperwork

Reduction Act of 1981 (44 USC, Chapter

35), the Peace Corps is requesting

emergency approval and clearance from

the Office of Management and Budget

for use of the Peace Corps Day

Brochure/Form to be used by the World

Wise Schools program. A copy of the

information collection may be obtained

from Monica Fitzgerald, Office of World

Wise Schools, Peace Corps, 1990 K St.,

NW, Washington, DC 20525. Ms.

Fitzgerald may be called at (202) 606–

9498. Peace Corps invites comments on

whether the proposed collection of

information is necessary for proper

performance of the functions of the

Peace Corps, including whether the

information will have practical use; the

accuracy of the agency’s estimate of the

burden of the proposed collection of

information, including the validity of

the methodology and assumptions used;

ways to enhance the quality, utility and

clarity of the information to be

collected; and ways to minimize the

burden of the collection of information

on those who are to respond, including

through the use of automated collection

techniques, when appropriate, and other

forms of information technology.

Comments on this form should be

addressed to Victoria Becker Wassmer,

Desk Officer, Office of Management and

Budget, NEOB, Washington, DC 20503.

Information Collection Abstract

Title: Peace Corps Day Brochure/

Form.

Need for and use of the Information:

This form is completed voluntarily by

Returned Peace Corps Volunteers and

educators throughout the country. This

information will be used by WWS to

identify individuals interested in

participating in the Peace Corps’s

annual Peace Corps Day program.

Enrollment in this program also fulfills

the third goal of Peace Corps as required

by Congressional legislation and to

enhance the Office of World Wise

Schools global education program.

Respondents: Returned Peace Corps

Volunteers and educators throughout

the public and private school systems in

the United States.

Respondents obligation to reply:

Voluntary.

Burden on the Public:

a. Annual reporting burden: 4,750 hrs.

b. Annual record keeping burden: 0

hrs.

c. Estimated average burden per

response: 3 min.

d. Frequency of response: annually.

e. Estimated number of likely

respondents: 95,000.

f. Estimated cost to respondents:

$0.79.

This notice is issued in Washington, DC on

January 23, 1998.

Bessy Kong,

Acting Associate Director for Management.

[FR Doc. 98–2020 Filed 1–27–98; 8:45 am]

BILLING CODE 6051–01–M

PENSION BENEFIT GUARANTY

CORPORATION

Submission of Information Collection

for OMB Review; Comment Request;

Allocating Unfunded Vested Benefits

AGENCY: Pension Benefit Guaranty

Corporation.

ACTION: Notice of request for extension

of OMB approval.